Technical Specification 3.7.10, Ultimate Heat Sink (UHS)

Indian Point 3 Improved Technical Specifications (ITS)

Conversion Package

Technical Specification 3.7.10:

"Ultimate Heat Sink (UHS)"

PART 1:

Indian Point 3 Improved Technical Specifications and Bases

Indian Point 3 ITS Submittal, Revision I

UHS 3.7.10

3.7 PLANT SYSTEMS

3.7.10 Ultimate Heat Sink (UHS)

LCO 3.7.10 The UHS shall be OPERABLE.

APPLICABILITY:

ACTIONS

MODES 1. 2, 3, and 4.

CONDITION REQUIRED ACTION COMPLETION TIME

A. UHS temperature > 950F. A.1 Be in MODE 3. 7 hours

OR AND

UHS inoperable for A.2 Be in MODE 5. 37 hours reasons other than temperature > 950F.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.7.10.1 Verify average water temperature of 24 hours UHS is • 95oF.

INDIAN POINT 3 Amendment [Rev.1], 06/28/003.7.10-1

UHS B 3.7.10

B 3.7 PLANT SYSTEMS

B 3.7.10 Ultimate Heat Sink (UHS)

BASES

BACKGROUND The UHS provides a heat sink for processing and operating heat from safety related components during a transient or accident, as well as during normal operation. This is done by utilizing the Service Water System (SWS) and the Component Cooling Water (CCW) System.

The ultimate heat sink for IP3 is the Hudson River. The UHS and supporting structures are capable of providing sufficient cooling for thirty days and are sufficient to:

(a) Support simultaneous safe shutdown and cooldown of both operating nuclear units at the Indian Point site and maintain them in a safe condition, and

(b) In the event of an accident in one unit, support required response to that accident and permit simultaneous safe shutdown and cooldown of the remaining unit and maintain them in a safe shutdown condition.

The ultimate heat sink is capable of withstanding the effects of the most severe natural phenomena associated with the Indian Point site, other site related events and a single failure of man-made structural features.

The two principal functions of the UHS are the dissipation of residual heat after reactor shutdown, and dissipation of residual heat after an accident.

APPLICABLE SAFETY- ANALYSES

The UHS is the sink for heat removed from the reactor core following all accidents and anticipated operational occurrences in which the unit is cooled down and placed on residual heat removal (RHR) operation. Because IP3 uses the UHS

(continued)

Revision [Rev.1], 06/30/00INDIAN POINT 3 B 3.7. 10 - 1

UHS B 3.7.10

BASES

APPLICABLE SAFETY ANALYSES (continued)

as the normal heat sink for condenser cooling via the Circulating Water System, unit operation at full power is its maximum heat load. Its maximum post accident heat load occurs shortly after a design basis loss of coolant accident (LOCA). Near this time, the unit switches from injection to recirculation and the containment cooling systems and containment recirculation system are required to remove the core decay heat.

The operating limits are based on conservative heat transfer analyses for the worst case LOCA. Reference 1 provides the details of the assumptions used in the analysis, which include worst expected meteorological conditions, conservative uncertainties when calculating decay heat, and worst case single active failure (e.g., single failure of a manmade structure). The UHS meets Regulatory Guide 1.27 (Ref.3), which requires a 30 day supply of cooling water in the UHS.

The UHS satisfies Criterion 3 of 10 CFR 50.36.

LCO The UHS is required to be OPERABLE and is considered OPERABLE if it contains water at or below the maximum temperature that would allow the SWS to operate for at least 30 days following the design basis LOCA without the loss of net positive suction head (NPSH), and without exceeding the maximum design temperature of the equipment served by the SWS. To meet this condition, the UHS temperature must not exceed 95oF.

APPLICABILITY In MODES 1, 2, 3, and 4, the UHS is required to support the OPERABILITY of the equipment serviced by the UHS and required to be OPERABLE in these MODES.

In MODE 5 or 6, the OPERABILITY requirements of the UHS are determined by the systems it supports.

(continued)

Revision [Rev.1], 06/30/00INDIAN POINT 3 B 3.7.10- 2

UHS B 3.7.10

BASES

ACTIONS A.1 and A.2

If UHS temperature > 95°F, or is inoperable for reasons other than high temperature, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 7 hours and in MODE 5 within 37 hours.

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.


This SR verifies that the SWS is available to cool the CCW System to at least its maximum design temperature with the maximum accident or normal design heat loads for 30 days following a Design-Basis Accident. The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. This SR verifies that the average water temperature of the UHS is - 950F. Requirements for UHS monitoring instrumentation are governed by the Technical Requirements Manual (Ref. 4). 1',

-IA -1

REFERENCES 1. FSAR, Section 9.6.

2. WCAP-12313, "Safety Evaluation For An Ultimate Heat Sink Temperature Increase To 95oF At Indian Point Unit 3"

3. Regulatory Guide 1.27.

4. IP3 Technical Requirements Manual.

INDIAN POINT 3 Revision [Rev.1]., 06/30/00B 3.7.10-3


Conversion Package

Technical Specification 3.7.10: "ULTIMATE HEAT SINK"

PART 2:

CURRENT TECHNICAL SPECIFICATION PAGES

Annotated to show differences between CTS and ITS

Indian Point 3 ITS Submittal, Revision 1

CTS AMENDMENT FOR AMENDMENT FOR COMMENT PAGE REV O SUBMITTAL REV 1 SUBMITTAL 3.3-10 145 145

3.3-10a 98 98 3.3-19 145;97-175 145;9-22-98 Change to Bases Page. No impact to ITS

3.7.10 T4.1-1 (5) 169;98-043 185 No impact T4.1-1 (6) 181 ;98-043 185 No impact

ITS 3.7.10

3. it the Component GooJlng System is not restored to meet the "requirements of 3.3.E.1 within the time periods specified in 3.3.E.2, then:

ITS L.7.&

a. If the reactor is critical, it shall be in the hot shutdown condition within four hours and in the cold shutdown condition within the following 24 hours.

b. If the reactor is subcritical, the reactor coolant system temperature and pressure shall not be increased more than 250F and 100 psi, respectively, over existing values. If the requirements of 3.3.E.1 are not satisfied within an additional 48 hours, the reactor shall be brought to the cold shutdown condition utilizing normal operating procedures. The shutdown shall start no later than the 9nd of the 48 hour period.LCO -T. Service wate s/te Ultmate Heat Sink' S3 .9. ' • . -

a. Three service water pumps on the cesignadc essential header and a minimum of two service water pumps on the designated Ir '.7% ~non-essential header, together with their associated piping & and valves, are operable.

0ý1 .I. b. _The service water inlet temperature is less than or equal to

S2. When the reactor is above cold shutdown and if the requirements of 3.3.F.l.a cannot be met within twelve hours, the reactor shall be brought to the cold shutdown condition, starting no later than the end of the twelve hour period, utilizing normal operating procedures.

L 3,-7.10 -3-. (When the reactor is ab. I . .As. down and if the requirement of (3.3.F.l.b is exceeded, the reactor shall be placed in at least hot shutdown within eve hours. and in cold shutdown within Aeh hours ýless the service water inlet temperature

decreases to wi he ent of 3.3.F.l.b.

3.3-10

Amendment No. ;#, Y$, 145

,4\

ITS 3.7.10

4. Isolation shall be maintained between the essential and non-essential headers at - all times when above cold shutdown conditions except that for a period of (5 3,7.9. -eight hours the headers may be connected while another essential header is being placed in service as described in F.2 above.

5. At least wo service ter inlet temperature monitorin instruments (a combination of in alled or port le instruments shall be operable en the reacto is above 35 F and service w er inlet temp ature exceeds OF.

6. 1 the requireme ts of 3.3.F.5 can t be met, the eactor shall be pfaced in t hot shutdown

condition w thin the next even hours and subsequentli cooled below O°F using normal operating ocedures.

7. Service water inlet temper ure shall be the ave ge of t or more servic water inlet tempe ture mon* oring instrument eadings per 3.3.F. taken wi in a five minute terval (instantaneo ).

8. When the reactor s above 350OF and rvice water inlet tem'eraturp/per 3.3.F.7 exceeds 0OF , service water inlet tenpoerature monitoring all commence at a frequency of" once per hour.

3.3-10 a

Amendment No. 98

/

A total of six service wate umps are installed. Only two of the set of thre service water pumps on header designated the essential header are required immediately followi a postulated loss-of-coolant accident. '8 During the recirculation phaoe-of the accident, two service water pumps on the non-essential header will beininually started to supply cooling water for one component cooling system hew exchanger, one control room air conditioner, and one diesel generatop< the other component cooling system heat exchanger, the other control roomair conditioner, the two other diesel generators and remai safety r ated equipment are cooled by the essential service water header During the .orecirculation phase of the acccident, both control room air co ioner units may be cooled by the essential service water header. The operability requirements on service wat temperature monitoring instrumentation and the frequency of service wate emperature monitoring insures that appropriate action can be taken to pre de operation beyond established limits. The locations selected for mo oring river water temperature are typically at the circulating or servic ater inlets, at the circulating water inlet boxes to the condenser hotwel or at the service water supply header to the fan cooler units. Temperat measurements at each of these locations are representative of the river wa r temperature supplied to cool plant heat loads.* Alternate locations may be cceptable on this basis. The limit on the service water maximum inlet te rature insures that the service water and component cooling water sysiems l.l be able to dissipate the heat loads generated in the limiting design ba s accident,.). This restriction allows up to seven hours for river water tempmrature transients which may temporarily increase the service water inlet mperature due to tidal effects to dissipate.

Tahse °srerelg ty of the equipment and systems required for the control of oren gas .sures that qupment is available to maintain hydrogen concentration within containment below the flammable limit ing post-LOCi 'conditions. Hydrogen concentration exceeding the fla le limit could potentially result in a containment wide hydrogen bu This could lead to overpressurization of containment, a breach of CONTAINWTGRITY, containment leakage, unacceptably high offsite doses, and to safety-related equipment located in containment. Two full rated rec ner units are provided in order ato control the hydrogen evolved in cont nment following a loss-of-coolant accident. Each unit is capable of pre ting the hydrogen concentration ,. om exceeding the flammable limit. ach recombiner is installed suck- t t independence is maintained and r dancy is assured. Each hydrogeeombiner S System C°dSonsi tsfalrecombin located inside containment, and eparate power supply, and control pan located outside containment s;d that they are accessible following esign basis accident.

3.3-19

Amendment No. $0, 00, X00, Z%, X4%, Revised by letter dated 9/22/98

TABLE 4.1-1 (Sheet 5 of 6)

* 1- * 1

Channel Description

37. Core Exit Thermocouples

38. Overpressure (OPS)

Protection System

42. Ambient Temperature Sensors Within the Containment Building

D

D

N.A.

N.A.

D

D

24M

18M (1)

N.A.

N.A.

24M

24M

Test

N.A.

18M

TM(1)

24M(2)

(1)

24M (2)

24M(3)

N.A.

N.A.

1) Calibration frequency for OPS sensors (RCS pressure and temperature) is 24 months

1) Independent operation of undervoltage and shunt trip attachments

2) Independent operation of undervoltage and shunt trip from Control Room manual push-button

1) Manual shunt trip prior to each use


3) Automatic undervoltage trip

IL I- I 4. 44.-

43. (Rjiver water mperature #t (instal

44. River Water T peraturee ## (portable%

s (1) Q (2)

I I I

10. A. 17 e against talle "instrumenta .n or another portabi evice

2) ibrate within days prior to use A quart y thereafter

-r�--� I� I I I I* *II. -

45. Steam Line Flow S 24M Q

_____________ I _____________ -

Engineered Safety Features circuits only

Amendment No. 7 9, 0, , 70, go, go, X07, Z28, 120, Z37, U0, 102, 104, 100, 6q

Ti

(2b

r(In

11

q

39. Reactor Trip Breakers

40. Reactor Trip Bypass Breakers

41. Reactor Vessel Level Indication System (RVLIS)

I

I

seZ0.

S. (Sheet 5 of 6)

Channel Description

37. Core Exit Thermocouples

38. Overpressure Protection System (OPS)

39. Reactor Trip Breakers

40. Reactor Trip Bypass Breakers

41. Reactor Vessel Level Indication System (RVLIS)

42. Ambient Temperature Sensors Within the Containment Building

43. River Water Temperature I (installed)

44. River Water Temperature 0 (portable)

45. Steam Line Flow

V T I V

hok

N.A.

N.A.

D

D

S

S (1)

S

Calibrate

24M

18H (1)

N.A.

N.A.

24M

24M

1814

0 (2)

24M

N.A.

24M

TM (1)

24M(2)

(1)

24M(2)

24M(3)

N. A.

N.A.

N.A.

N.A.

Q

I i -

Remarks

1) Calibration frequency for OPS sensors (RCS pressure and temperature) is 24 months

1) Independent operation of undervoltage and shunt trip attachments


1) Manual shunt trip prior to each use


3) Automatic undervoltage trip

1) Check against installed instrumentation or another portable device

2) Calibrate within 30 days prior to use and quarterly thereafter

Engineered Safety Features circuits onl v

Amendment No. 33, It, 05, 74, 70, 93, 03, •07, U2S, 120, 137,

(x�

I

I

too, 192, Ifit, M•, 185

TABLE L.j (Sheet 6 of 6) ITS 3.7.10

Table Notation

A By means of the movable incore detector system

** Quarterly when reactor power is beiow the setpoint and prior to each startup if not done previous month.

8EE

# T hese rega ements e applic when spelr ion' V 3...I is in X_ .-. 7/. • eff -only.

S## The "each shift" frequency also requires veritication that the ::N, parameters (Reactor Coolant Temperature, Reactor Coolant Flow, and Pressurizer Pressure) are within the limits of Technical Specifization 3.1.H.

S - Each Shift

GEE W - Weekly P - Prior to each startup if not done previous week CI7S M - Monthly

11 TSFO NA -- Not Applicable Q - Quarterly D - Daily 18M - At least once per 18 months TM - At least every two months on a staggered test basis (i.e., one train

per month) 24M - At least once per 24 months 6M - At least once per 6 months

Amendment No. X27, ZX5, 1X07, Z OO , 0, 170,

I �-

B (Sheet 6 of 6)

Table Notation

"By means of the movable incore detector system ** Quarterly when reactor power is below the setpoint and prior to

each startup if not done previous month.

These requirements are applicable when specification 3.3.F.5 is in effect only.

## The "each shift" frequency also requires verification that the DNB parameters (Reactor Coolant Temperature, Reactor Coolant Flow, and Pressurizer Pressure) are within the limits of Technical Specification 3.1.H.

S - Each Shift (i.e.,at least once per 12 hours) W - Weekly P - Prior to each startup if not done previous w M - Monthly NA - Not Applicable Q - Quarterly D - Daily 18M - At least once per 18 months TM - At least every two months on a staggered tes'

per month) 24M - At least once per 24 months 6M - At least once per 6 months

eek

t basis (i.e., one train

Amendment No. 137, ZS4, 107, ZOO, 199, 170, Z1Z, 185


Conversion Package


"Ultimate Heat Sink (UHS)"

PART 3:

DISCUSSION OF CHANGES

Differences between CTS and ITS


DISCUSSION OF CHANGES ITS SECTION 3.7.10 - Ultimate Heat Sink (UHS)

ADMINISTRATIVE

A.1 In the conversion of the Indian Point Unit 3 Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS) certain wording preferences or conventions are adopted which do not result in technical changes (either actual or interpretational). Additionally, editorial changes, reformatting, and revised numbering are adopted to make ITS consistent with the conventions in NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Rev. 1, i.e., the Improved Standard Technical Speci fi cati ons.

The CTS Bases are deleted and replaced with comprehensive ITS Bases designed to support interpretation and implementation of the associated Technical Specifications. The Bases explain, clarify, and document the reasons (i.e., bases) for the associated Technical Specifications, and reflect the IP3 plant specific design, analyses, and licensing basis. In accordance with 10 CFR 50.36(a), the ITS Bases are included with the proposed ITS conversion application; however, deletion of the CTS Bases and the adoption of the ITS Bases is an administrative change with no impact on safety.

A.2 CTS Limiting Conditions for Operation (LCOs) and Surveillance Requirements (SRs) include statements of the objective and the applicability. The CTS statements of objective and applicability are deleted because these statements do not establish any requirements and do not provide any guidance for the application of CTS requirements. Therefore, deletion of these statements has no significant adverse impact on safety.

A.3 CTS 3.3.F specifies the Applicability for Service Water System/Ultimate Heat Sink as whenever the reactor is above cold shutdown (i.e., Modes 1, 2, 3 and 4). ITS 3.7.9 and ITS LCO 3.7.10 maintain this Applicability by requiring that the Service Water System and Ultimate Heat Sink be Operable in Modes 1, 2, 3 and 4. This is an administrative change with no impact on safety because there is no change to the CTS Applicability.

ITS Conversion Submittal, Rev 1Indian Point 3 I

DISCUSSION OF CHANGES ITS SECTION 3.7.10 - Ultimate Hea, Sink IUHS)

A.4 CTS 3.3.F.3 requires that the reactor be placed at least hot shutdown (i.e., Mode 3) within 7 h•rs, and in cold-shutdow'r (i.e., Mode 5) "within the.following 30 hou . if the service.water\.nlet temperature ) exceeds 95°F. Under the same ondition, ITS LCO 3.7.1 Required Actions B.1 anXB.2 require that the re tor be placed in Mode within 6 hours and in e 5 within 36 hours. T is change is needed bec use it makes K the plant utdown Completion Times onsistent with other NS LCOs. This change acceptable because the TS Completion Times \]low sufficient tim to reach the required u it conditions from fuli power conditions in an rderly manner and witho t challenging unit systems. This is an adminis ative change with no i , act on safety because",there is no significant ch e to the existing req rements. ",

S.5 CTS 3.F.l.b requires thaUHS (service water in\_t) temperature is (

less t n 950F. The CTS Base for CTS 3.3.F.1.b cla ifies that this restrict n is intended to all up to seven hours fo, the dissipation of tidal e ects that can cause iver water temperatur transients that ( may temporarily increase UHS temperture. "ITS LCO 3.7.10, ndition A and Requred Action A1 maintai the

Req,,e ActionqA.1, " in • e •requirement'that uibe maintained less'than 950F. Additionally, the

quired Action A.1 ompletion Time of 7 urs for restoration f UHS tem rature before a r actor shutdown mustxe initiated provides. I "exli t recognition of he allowance for th dissipation of tidal effects hat can cause ri r water temperature ransients that may ( temporari increase UHS te erature. This is a administrative change with no adv se impact of sa ty because it is an explicit statement of "an existing C pletion Time st ed in the CTS Bases..

MORE RESTRICTIVE

M.1 CTS 3.3.F.1.b requires that UHS (service water inlet) temperature is less than 95°F: however, there is no requirement for periodic verification that this limit is met except that CTS 3.3.F.8 includes a requirement to monitor UHS temperature every hour when the UHS temperature exceeds 900F. ITS SR 3.7.10.1 requires verification every 24 that the average temperature of the Ultimate Heat Sink is _< 951F regardless of the UHS temperature; and, requirements for accelerated

Indian Point 3 2 TT, rnnwarcinn Q,,krni-÷+l D 1.1. . '', , ..J •J U L I U I ,J LAJU III I M L , I , t' 1 V I


verification of UHS temperature when UHS temperature is approaching the upper limit are relocated to the Technical-Requirements Manual (TRM)

ý-,:..:(See ITS 3.7.10, DOC LA.1). This more ,restri~ctive requirement is acceptable because it verifies that the UHS temperature is consistent with the accident analysis assumptions, and the 24 hour Frequency is acceptable based on operating experience related to trending of the parameter variations during the applicable modes. This change has no adverse impact on safety.

LESS RESTRICTIVE

None

REMOVED DETAIL

LA.1 CTS 3.3.F.5 through CTS 3.3.F.8 require accelerated monitoring (once per hour) using specific instruments whenever the UHS temperature is > 90OF (i.e., approaching the LCO limit of 951F). CTS Table 4.1-1, Items 43 and 44 require periodic channel checks and calibrations of the instruments used to perform these verifications.

ITS LCO 3.7.10 maintains the requirement that UHS be maintained less than 950F and ITS SR 3.7.10.1 maintains the requirement for verification every 24 hours that this limit is met; however, requirements for accelerated monitoring of UHS temperature using specific instruments whenever the UHS temperature is approaching the LCO limit of 951F are relocated to the TRM.

This change is acceptable because ITS LCO 3.7.10 maintains the requirement that the LCO is met whenever the plant is in the Applicable Modes. Maintaining this requirement in Technical Specifications and maintaining requirements for accelerated monitoring of UHS temperature using specific instruments whenever the UHS temperature is approaching the LCO limit provides an adequate level of assurance of prompt identification and initiation of actions if UHS temperature exceeds required limits.

The Quality Assurance Plan will be revised to specify that requirements

in the TRM are part of the facility as described in the FSAR and that

Indian Point 3 3 TT, rnn\rcinn QKm i+ I D, 1'-,'-,,J'-, ,JI U M-. . - aJ L IIII ,U I I , t:7V I

v


changes to the TRM can be made only in accordance with the requirements of 10 CFR 50.59. Therefore, this change is acceptable because there is no change to the existing requirements by the relocation of requirements to the TRM and future changes to the TRM will be controlled in accordance with 10 CFR 50.59.

This change is a less restrictive administrative change with no impact on safety because ITS 3.7.10 maintains the requirement that plant operation be curtailed if UHS temperature exceeds the upper limit. Therefore, requirements for accelerated monitoring of UHS temperature and the instruments used to perform this accelerated monitoring can be maintained in the TRM with no significant adverse impact on safety.

ITS Conversion Submittal, Rev 1Indian Point 3 4


Conversion Package

Technical Specification 3.7.10: "Ultimate Heat Sink (UHS)"

PART 4:

No Significant Hazards Considerations for

Changes between CTS and ITS that are

Less Restrictive No Significant Hazard Considerations for Changes that are Administrative, More Restrictive, and Removed Details are the same for all Packages. A Copy is included at the end of the Package.


NO SIGNIFICANT HAZARDS EVALUATION ITS SECTION 3.7.10 - Ultimate Heat Sink (UHS)

LESS RESTRICTIVE

("L.1" Labeled Comments/Discussions)

There are no less restrictive changes for the adoption of this ITS.



Conversion Package

Technical Specification 3.7.10: "Ultimate Heat Sink (UHS)"

PART 5:

NUREG-1431 Annotated to show differences between

NUREG-1431 and ITS

Status of NUREG 1431 Generic Changes for ITS 3.7.10 This ITS Specification is based on NUREG-1431 Specification No. 3.7.9 as modified by the following Generic Changes:

OG No. TSTF No. Generic Change Description NRC STATUS IP3 STATUS JD No.

N/A NIA NO GENERIC CHANGES ARE Not Applicable Not Applicable NIA POSTED AGAINST THIS SPECI FI CATI ON.


UHS 3.7.Y 10

3.7 PLANT SYSTEMS

S3F. 1->

3.7.4 Ultimate J0

LCO 3.7.g I0

Heat Sink (UHS)

The UHS shall be OPERABLE.



SR .'.9.1/Verify/water •yvel ofAHS is-/562]fe hoursi/

[mea sea Ie]. /4ct n/

(continued)

WOG ST e ,0/79

t ! ! 1 I

!

E-L DG,

WOG STS Rev 1, 04/07/95

NUREG-1431 Markup Inserts ITS SECTION 3.7.10 - Ultimate Heat Sink (UHS)

INSERT: 3.7-21-01

UHS temperature > 951F.

INSERT: 3.7- 21-02

Verify UHS te erature O 0F.

UHS 3. 7S.,1j)

SURVEILLANCE REQUIREMENTS (continued)


.37ý.2 uVerify avera e water temperature of 24 hours F ,1')UNS is 5 r tF.J KoS M-s,-rit,ý

Rev 1, 04/07/95WOG STS 3.7-22

UHS B 3.7.y )b

8 3.7 PLANT SYSTEMS

B 3.7.Y Ultimate Heat Sink (U14S)

BASES

The UHS provides a heat sink for processing and operating heat from safety related components during a transient or

accident, as well as during normal operation. This is done

by utilizing the Service Water System (SWS) and the

Component Cooling Water (CCW) System.

inclZu- -gnecess y retainingstructuresje'g., a pon"wl its am, or a ver with it dam), and -.e canals 0 cgiuits con cting the s rces with ut not in ding,

-oling wa r system in ke struct es as disc ed in t.

FSAR, Se on [9.z.53 Ref. 1). cooling t ers or /

portio thereof ar required accomplis he4 UMS sael funct, ns t s dme

in . The two principal functions of the UHS are the

dissipation of residual heat after reactor shutdown, and

dissipation of residual heat after an accident.

Avariety compl es is used to meet thi aUKS. )'lake oran oan n mya qualify as

if the omplex )icludes an/ter source cg

stru ure, it ys likely at a second

(continued)

Rev 1, 04/07/95

FQ

BACKGROUND

WOG STS

{


INSERT: B 3.7-46-01

The ultimate heat sink for IP3 is the Hudson River. The UHS and supporting structures are capable of providing sufficient cooling for thirty days and are sufficient to:

(a) Support simultaneous safe shutdown and cooldown of both operating nuclear units at the Indian Point site and maintain them in a safe condition, and

(b) In the event of an accident in one unit, support required response to that accident and permit simultaneous safe shutdown and cooldown of the remaining unit and maintain them in a safe shutdown condition.

The ultimate heat sink is capable of withstanding the effects of the most severe natural phenomena associated with the Indian Point site, other site related events and a single failure of man-made structural features.

UHS B 3.7.9

BASES

BACKGROUND Ad frmat _ on desig Wan ra n o (continued) sys a listof comprn ents s rve can oun i••ef ence ýelnn Ig

APPLICABLE The UHS is the sink for heat removed from the reactor core SAFETY ANALYSES following all accidents and anticipated operational

occurrences in which the unit is cooled down and placed on residual heat removal (RHR) operation. or ( ff itt -

- UHS as the normal heat sink for condenser cooling via the Circulating Water System, unit operation at full power is its maximum heat load. Its max'imum post accident heat load _occur after a design basis loss of coolant

••A-"•accident (LOCA). Near this time, the unit switches from injection to recirculation and the containment cooling systems and are required to remove the core decay heat.

Co-•djJ( The operating limits are based on conservative heat transfer analyses for the worst case LOCA. Reference 1 provides the details of the assumptions used in the analysis, which include worst expected meteorological conditions, conservative uncertainties when calculating decay heat, and worst case single active failure (e.g., single failure of a manmade structure)). he UHS • ed * I. 25r2da n-wiI Regulatory Guide 1.27 (Ref. ',1 which requires a 30 day supply of cooling water in the

The UHS satisfies Criterion 3 of (,{hhfM 'licyntatemenJD

LC'0 The UHS is required to be OPERABLE and is considered OPERABLE if it contains _-__•__ i nMn. water at or below the maximum temperature that would allow the SWS to operate for at least 30 days following the design basis LOCA without the loss of net positive suction head (NPSH), and without exceeding the maximum design temperature of the equipment served by the SWSý. To meet~this co-ndition- Vie

G ;ý-- • --- U't~emerat uree not exceed r90'n ndt~he 54et___ m uhudntTa,, I ow E5'H-ft mea-n- s-7-evell during no-rma

(continued)

ýR5

WOG STS B 3.7-47 Rev 1, 04/07/95-

UHS B 3.7.9

BASES (continued)

APPLICABILITY In MODES 1, 2, 3, and 4, the UHS is required to support the OPERABILITY of the equipment serviced by the UHS and required to be OPERABLE in these MODES.

In MODE 5 or 6, the OPERABILITY requirements of the UHS are determined by the systems it supports.

ACTIONS'If one or more~xo l n 1!er 1h1alve1 one 11n111ope11H e (i.e.,•

Sup to one f~rper cooin owrinoperable), a on must be]

t take to ;;etore the tnoper Y. e cooling towagefan(s) to | >1 OPEIALE/status within 71as

X h~ day Completio /ime is reason e based on the low/

one cooling tow fan is oper e (in one or more. ooling towers), the mber of avail e stems, and th. i-e required t reasonably cop ete the Required. ion.

N,

S•L• • • itin týassociated Co*,letionlior if-'the UHS) ~~* lo• no eIble for nasons other tli oniti A theunt•

Si~mst be placed in a MuuL in wnic-l te LCO dentapply. S• / To achieve this status, the unit must be placed in at least

MOD 3 within 6 hours and in 140DE 5 within 36 hours,

- ertg-xr ien e, t I re the rqinr aed ni.t easdtl

I frmfl oe odtos in an orderl y manner and wi thout Schallenging unit systems. •


(continued)

Rev 1, 04/07/95

/J

R 3. /

Th SR ver* ies that dequate lo term (3 day) cooli n be ma •tained. e specifi level a o ensures at

s fici t NPSH i available operate e SWS p The / [24] ur Freque y is base on opera ng experien relatdr

to ending of he parame r variations during t h

f

B 3.7-4BWO STS


INSERT: B 3.7-48-01 or i-v inop o {w Vll-c, to•]v t -v\

If UHS temperature > 950F the-U S temperature'muqst be verified to be <__5T withi.n hours. The 7 hour Completion Time6llows for the, (diss'-ation of ial effects that can ause river wat- temperature

ansie that may orarily increase calized UHS tomperature.

INSERT: B>3\-48-02

UHS temper ture does no return to • 95OF

P!

UHS B 3.7.9

BASES

SURVEILLANCE SR 3.7.9.1 (continued) REQUIREMENTS REQUREMNTS applicable MODES. This SR verifies that the UHS water level|

L is k [562] ft [mean sea level].

SR 3.71:6 100

This SR verifies that the SWS is available to cool the CCW System to at least its maximum design temperature with the maximum accident or normal design heat loads for 30 days following a Design Basis Accident. The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. This SR verifies that the average water temperature of the UHS is

L :5

SR 2(7.9.3

eratin each c ling to er fan for [15] mi es ensures that al fans a OPERA E and tha all assoc*ted control are f ctioni proper . It als ensures t at fan or ;o or fail re, or cessiv :vibration can be de cted for / co ective ction. he 31 da• requency Is based on/ o terating experie e, the kn reliabi ty of the f$ nit redu ancy avail le, and e low probabIl ity of signifcant de adation o the UHS c ling tower fins occu ing be een survei ances. / /

3.7. . / - / This verifies hat each ooling tower fan starts nd oper es on an ctual or ,/mulated act tion signa . The [1 month Fr quency is onsistent w h the typi; r ueling c le. Operaing experi !ce has showif that the omponent usually pas the Survei lance when 4erformed the [18] onth Frequency. There ore, the Fr quency is

Saccepta le from a reliability standpoint.

REFERENCES 1. FSAR, Section 0 01.'(9,'21 /OG.TRegulatory Guide 1.77.

WOG STS B 3.7-49 Rev 1, 04/07/95


INSERT: B 3.7-49-01

Requirements for UHS monitoring instrumentation are governed by the Technical Requirements Manual (Ref. 4)

INSERT: B 3.7-49-02

2. WCAP-12313, "Safety Evaluation For An Ultimate Heat Sink Temperature Increase To 95°F At Indian Point Unit 3"

INSERT: B 3.7-49-03


I

I


Conversion Package

Technical Specification 3.7.10: "Ultimate Heat Sink (UHS)"'

PART 6:

Justification of Differences between

NUREG-1431 and IP3 ITS


JUSTIFICATION OF DIFFERENCES FROM NUREG-1431 ITS SECTION 3.7.10 - Ultimate Heat Sink (UHS)

RETENTION OF EXISTING REQUIREMENT (CURRENT LICENSING BASIS)

CLB.1 NUREG-1431, Rev 1, Section 3.7.10, was modified as needed to reflect the IP3 design and current licensing basis. A detailed description of the design, accident analysis assumptions, and Operability requirements are incorporated into the IP3 ITS Bases. These changes maintain the IP3 current licensing basis except as identified and justified in the CTS/ITS discussion of changes.

PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT

PA.1 Corrected typographical error or made a minor editorial improvement to improve clarity and ensure requirements are fully understood and consistently applied. There are no technical changes to requirements as specified in NUREG 1431, Rev. 1: therefore, this change is not a significant or generic deviation from NUREG 1431, Rev 1.

PLANT-SPECIFIC DIFFERENCE IN THE DESIGN OR DESIGN BASIS

DB.1 Design or implementation details are incorporated or revised as necessary to more precisely describe IP3 current design or practice. These changes are intended to describe the design, improve clarity, or ensure requirements are fully understood and consistently applied. Unless identified and described klow, these changes are selfexplanatory. A detailed description of the design, accident analysis assumptions, and Operability requirements are incorporated into the IP3 ITS Bases. These changes maintain the IP3 current licensing basis except as identified and justified in the CTS/ITS discussion of changes. There are-no technical changes to requirements as specified in NUREG 1431, Rev 1; therefore, this change is not a significant or generic deviation from NUREG 1431, Rev 1.

DIFFERENCE BASED ON A GENERIC CHANGE TRAVELER FOR NUREG-1431

None


JUSTIFICATION OF DIFFERENCES FROM NUREG-1431 ITS SECTION 3.7.10 - Ultimate Heat Sink (UHS)

DIFFERENCE FOR ANY REASON OTHER THAN ABOVE

None



Conversion Package

Technical Specification 3.7.11: "Control Room Ventilation System (CRVS)"

PART 1:


N U IL)


'S ? Q- c, P 0% D Q-S

CRVS B 3.7.11

B 3.7 PLANT SYSTEMS

B 3.7.11 Control Room Ventilation System (CRVS)

BASES

BACKGROUND The CRVS provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity, chemicals, or toxic gas.

The Control Room Ventilation System consists of the following equipment: a single filter unit consisting of two roughing filters, two high efficiency particulate air (HEPA) filters: two activated charcoal adsorbers for removal of gaseous activity (principally iodines); two 100 capacity filter booster fans; and, a single duct system including dampers, controls and associated accessories to provide for three different air flow configurations. The air-conditioning units associated with the CRVS are governed by LCO 3.7.12, "Control Room Air Conditioning System (CRACS)."

The CRVS is divided into two trains with each train consisting of a filter booster fan and the associated inlet damper and the following components which are common to both trains: the control room filter unit, damper A (filter unit bypass for outside air makeup to the Control Room), damper B (filter unit inlet for outside air makeup to the Control Room), damper C (filter unit inlet for reticulated air), and the toilet and locker room exhaust fan. The two filter booster fans (F 31 and F 32) are powered from safeguards power trains 5A (EDG 33) and 6A (EDG 32), respectively. The automatic dampers that are common to both trains are positioned in the fail-safe position (open or closed) by either of the redundant actuation channels.

The CRVS is an emergency system, parts of which operate during 0, normal unit operations.

The three different CRVS air flow configurations are as follows:

(continued)

INDIAN POINT 3 Revision [Rev.1], 06/30/00B 3.7.11-1

CRVS B 3.7.11

BASES

BACKGROUND a) Normal operation consists of approximately 85% (8500 cfm) (continued) unfiltered recirculated flow driven by the air-conditioning fans and approximately 15% (1500 cfm) unfiltered outside air makeup;

b) Incident mode with outside air makeup (i.e. 10% incident mode) consists of approximately 87% (9250 cfm) unfiltered recirculated flow driven by the two safety related airconditioning fans, at least 10% (> 1000 cfm) filtered recirculated flow driven by either one of the two filter booster fans and approximately 2.5% to 4.0% (250 to 400 cfm) filtered outside air makeup;

c) Incident mode with no outside air makeup (i.e. 100% incident mode) consists of 85% (9100 cfm) unfiltered recirculated ..- I flow driven by the two safety related air-conditioning fans, approximately 15% filtered recirculated flow driven by 0 either one of the two filter booster fans and no outside air makeup.

Note that the required recirculation rates are demonstrated with surveillance tests conducted with the air conditioning system (CRACS) operating. An inoperable CRACS fan will affect the flow balance of the CRVS due to interconnected ductwork. Therefore, if the fan associated with one of the air-conditioning units governed by LCO 3.7.12 is inoperable, Conditions in both LCO 3.7.11, Control Room Ventilation System, and LCO 3.7.12, Control Room Air Conditioning System (CRACS), will apply.

Incident mode with outside air makeup is the preferred method of operation during any radiological event because it provides outside air for pressurization of the Control Room. Calculations indicate that very low volumes of outside air makeup will maintain the Control Room at a slight positive pressure. Nevertheless, due to the difficulty of adjusting and maintaining the flow dampers to provide a low flow, it was determined that the damper should be adjusted to provide a flow of approximately 250 cfm (2.5% outside air makeup). However, a higher volume of outside air makeup to

(continued)

INDIAN POINT 3 Revision [Rev.1], 06/30/00B 3.7.11 - 2

CRVS B 3.7.11

BASES

BACKGROUND the Control Room increase the thyroid dose to the operators during (continued) an accident. Therefore, the Control Room dose assessment assumes

a filtered outside air makeup of approximately 400 cfm (4.0% outside air makeup).

On a Safety Injection signal or high radiation in the Control Room (Radiation Monitor R-1), the CRVS will actuate to the incident mode with outside air makeup (i.e. 10% incident mode). This will cause one of the two filters booster fans to start, the locker room exhaust fan to stop, and CRVS dampers to open or close as necessary to filter all incoming outside air and direct approximately 10% of the recirculated air through the filter unit. In the event that the first booster fan fails to start, the second booster fan will start after a predetermined time delay.

If for any reason it is required or desired to operate with 100% recirculated air (e.g., toxic gas condition is identified), the CRVS can be placed in the incident mode with no outside air makeup (i.e. 100% incident mode) by remote manually operated switches. The Firestat detectors will also initiate 100% incident mode in the CRVS.

The control room is continuously monitored by radiation and toxic gas detectors. On a Safety Injection signal or high radiation in the Control Room (Radiation Monitor R-1), will cause actuation of the emergency radiation state of the CRVS (i.e., i with outside air makeup (i.e. 10% incident mode)).

The CRVS does not actuate automatically in response to toxic gases. Separate chlorine, ammonia and oxygen probes are provided to detect the presence of these gases in the outside air intake. Additionally, monitors in the Control Room will detect low oxygen levels and high levels of chlorine and ammonia. The CRVS may be placed in the incident mode with no outside air makeup (i.e. 100% incident mode) to respond to these conditions. Instrumentation for toxic gas monitoring is governed by the IP3 Technical Requirements Manual (TRM) ( Ref. 4). Generally, the manually initiated actions of the toxic gas isolation state are more restrictive, and will override the actions of the emergency radiation state.

(continued)

INDIAN POINT 3 R 1 7 11 - . .....* I * .LJ. - nersion LKev.ij, Ub/oU/uu

CRVS B 3.7.11

BASES

BACKGROUND A single train will create a slight-positive pressure in the (continued) control room. The CRVS operation in maintaining the control room

habitable is discussed in the FSAR, Section 9.9 (Ref. 1).

The CRVS is designed in accordance with Seismic Category I requi rements.

The CRVS is designed to maintain the control room environment for 30 days of continuous occupancy after a Design Basis Accident (DBA) without exceeding a 5 rem whole body dose or 30 rem to the thyroid.

APPLICABLE SAFETY ANALYSES

The CRVS active components are arranged in redundant, safety related ventilation trains. The location of components and ducting within the control building envelope ensures an adequate supply of filtered air to all areas requiring access. The CRVS provides airborne radiological protection for the control room operators, as demonstrated by the control room accident dose analyses for the most limiting design basis accident (i.e., DBA LOCA) fission product release presented in the FSAR, Chapter 14 (Ref. 2).

Radiation monitor R-1 is not required for the Operability of the Control Room Ventilation System because control room isolation is initiated by the safety injection signal in MODES 1, 2, 3, 4, and control room isolation is not required for maintaining radiation exposure within General Design Criteria 19 limits following a fuel handling accident or gas-decay-tank rupture.

The worst case active failure of a component of the CRVS, assuming a loss of offsite power, does not impair the ability of the system to perform its design function. However, the original CRVS design was not required to meet single failure criteria and, although upgraded from the original design, CRVS does not satisfy all requirements in IEEE-279 for single failure tolerance.

(continued)

INDIAN POINT 3 Revision [Rev.1], 06/30/00B 3.7.11 - 4

CRVS B 3.7.11

BASES


Each of the automatic dampers that are common to both trains is positioned in the fail-safe position (open or closed) by either of the redundant actuation channels.

The CRVS satisfies Criterion 3 of 10 CFR 50.36.

LCO Two CRVS trains are required to be OPERABLE to ensure that at least one is available. Total system failure could result in exceeding a dose of 5 rem whole body or 30 rem to the thyroid of the control room operator in the event of a large radioactive release.

The CRVS is considered OPERABLE when the individual components necessary to limit operator exposure are OPERABLE in both trains. A CRVS train is OPERABLE when the associated:

a. Filter booster fan and an air-conditioning unit fan powered from the same safeguards power train are OPERABLE;

b. HEPA filters and charcoal adsorbers are not excessively restricting flow, and are capable of performing their filtration functions; and

c. Ductwork, valves, and dampers are OPERABLE or in the incident mode, and air circulation can be maintained.

In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, ductwork, and access doors.

Instrumentation for toxic gas monitoring is governed by the IP3 Technical Requirements Manual (TRM) ( Ref. 4) and is not included in the LCO.

Note that the required recirculation rates are demonstrated with surveillance tests conducted with the air conditioning system (CRACS) operating. An inoperable CRACS fan will affect the flow

(continued)

INDIAN POINT 3 Revision [Rev.1], 06/30/00B 3.7.11- 5

CRVS B 3.7.11

BASES

LCO (continued)

APPLICABILITY

ACTIONS

balance of the CRVS due to interconnected ductwork. Therefore, if the fan associated with one of the air-conditioning units governed by LCO 3.7.12 is inoperable, Conditions in both LCO 3.7.11, Control Room Ventilation System, and LCO 3.7.12, Control Room Air Conditioning System (CRACS), will apply.

In MODES 1, 2, 3, 4 CRVS must be OPERABLE to limit operator exposure during and following a DBA.

The CRVS is not required in MODE 5 or 6, or during movement of irradiated fuel assemblies and core alterations because analysis indicates that isolation of the control room is not required for maintaining radiation exposure within acceptable limits following a fuel handling accident or gas decay tank rupture.

Administrative controls address the role of the CRVS in maintaining control room habitability following an event at Indian Point Unit 2.

A.1

When one CRVS train is inoperable, action must be taken to restore OPERABLE status within 7 days. In this Condition, the remaining OPERABLE CRVS train is adequate to perform the control room protection function. However, the overall reliability is reduced because a failure in the OPERABLE CRVS train could result in loss of CRVS function. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.

B.1

When neither CRVS train is Operable, action must be taken to restore at least one train to OPERABLE status within 72 hours. The 72 hour Completion Time is acceptable because of the low probability of a DBA occurring during this time period.

(continued)


CRVS B 3.7.11

BASES

ACTIONS C.1 and C,2 (continued)

If Required Actions A.1 or B.1 are not met within the required Completion Time, the unit must be placed in a MODE that minimizes accident risk. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.


Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing each train once every month provides an adequate check of this system. Note that a CRVS train includes both the filter booster fan and an airconditioning unit fan powered from the same safeguards power train. The 31 day Frequency is based on the reliability of the equipment and the two train redundancy availability.

This SR verifies that the required CRVS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The CRVS filter tests are in accordance with the sections of Regulatory Guide 1.52 (Ref. 3) identified in the VFTP. The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, minimum flow rate, and the physical properties of the activated charcoal. Specific test Frequencies and additional information are discussed in detail in the VFTP.

(continued)


CRVS B 3.7.11

BASES

SURVEILLANCE REQUIREMENTS (continued)

This SR verifies that each CRVS train starts and operates on an actual or simulated actuation signal. The Frequency of 24 months is based on operating experience which has demonstrated this Frequency provides a high degree of assurance that the booster fans will operate and dampers actuate to the correct position when required.

This SR verifies the integrity of the control room enclosure, and the assumed inleakage rates of the potentially contaminated air. The control room positive pressure, with respect to potentially contaminated adjacent areas, is periodically tested to verify proper functioning of the CRVS. During the operation in the incident mode with outside air makeup (i.e. 10% incident mode), the CRVS is designed to maintain the control room at a slight positive pressure with respect to adjacent areas in order to prevent unfiltered inleakage. The CRVS is designed to maintain this positive pressure with very low volumes of outside air makeup. Due to the difficulty of adjusting and maintaining the flow dampers to provide a low flow, it was determined that the damper should be adjusted to provide a flow of approximately 250 cfm (2.5% outside air makeup). Note that the higher the volume of outside air makeup to the Control Room, the higher the thyroid dose to the operators during an accident. The acceptance criteria of 400 cfm (4.0% outside air makeup) is the volume used in the Control Room dose assessment.

The SR Frequency of 24 months on a staggered test basis is acceptable because operating experience has demonstrated that the control room boundary is not normally disturbed. Staggered testing is acceptable because the SR is primarily a verification of Control Room integrity because fan operation is tested elsewhere.

(continued)


CRVS B 3.7.11

BASES (continued)


2. FSAR, Chapter 14.

3. Regulatory Guide 1.52, Rev. 2.




Conversion Package

Technical Specification 3.7.11: "Control Room Ventilation System (CRVS)"

PART 5:


NUREG-1431 and ITS


OG No. TSTF No. Generic Change Description NRC

BWOG-009 ADD AN ACTION FOR LOSS OF Rejec CONTROL ROOM ENCLOSURE INTEGRITY

BWROG-017 051 RO REVISE CONTAINMENT NRC REQUIREMENTS DURING HANDLING IRRADIATED FUEL AND CORE ALTERATIONS (REQUIREMENTS LIMITED TO "RECENTLY" IRRADIATED FUEL)

WOG-086 287 R2 VENTILATION SYSTEM NRC F ENVELOPE ALLOWED OUTAGE TIME

STATUS

ted by TSTF

IP3 STATUS

Not Incorporated

JD No.

NIA

Review Not Incorporated NIA

Review Not Incorporated NIA


B 3.7.X I'

B 3.7 PLANT SYSTEMS

B 3.7.1k, Control Room ) System -(GREF-S)

BASES

BACKGROUND The UEM-Eprovides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity4, chemicals, or toxic gasý.

8�7 5?-o'

KTI2X3 k 1.-

the CREFS onsist of two independent, redu ant trains t at recirc 0; aten and ilter the 6ntrol room . Each tra* cons• ts of refilter o0demister, a igh efficien p iculat ir (HEPA) ilter, an ac vated charco dsorber ection for emoval of ga ous activity (princi ally iodin ), and a fa Ductwork, v ves or am s, and in umentation o form part the sys m, a well as de *sters to re e water drop s from t air tream. A cond bank of PA filters lows th dsorber

section t collect carb fines and p vide back in case--. O? tail Ire or 'rne malw'Hj•?T1i iter DanK,. eevs The CREF is an emergency system, parts of which mya-el operate during normal unit operationsopa•On t ach tpei (• eain V UpJr -e t ofAhe tAc t u at in si•n-alT is, 6norp m

fir su y 10 th r controm ronts wisolate he e st o am reof ve milationuir il ed throPA the syste ilter - -tr*s. The d efilters a demi sters i ove any latt

ffecticles i the air, any entrais.d water droets ~~ .. /rese~nt, ; prevent •cessive loAd i g of the HE P filters /

• ]nd chla ~t asrr.C in mis operation each trln Sfr a a 0 r prmn with the he lers on, / r dr soitu ulupo EA fil ter• and adso bers.

(B th eie n et ae impoprtan to the/ ' 'fectivenes• of the charx~oal adsorbers./

Actuation of he CREFS pl ces the system in either of two\ separate ates (emerg y radiation s te or toxicais isolat' n state) ofe emergency mo i of operati depe ing on the tiation signal Actuation the system t he emergenc radiation stat of the emer cy mode of peration, cl es the unfilte d outside ai intake and

unfiltered aust dampers nd aligns th system for recircula on of the c room air t ough the r undant trains HEPA and the arcoal filte . The eme ency radi ion state also nitiates pres rlization a filtered ve ilation of the ir supply to e control r om.

. (r.nt4~wed~-)-

WOG~~~~~ .T 50Rv1 0/79

I

WOG STS Rev 1, 04/07/95

NUREG-1431 Markup Inserts ITS SECTION 3.7.11 - Control Room Ventilation System (CRVS)

INSERT: B 3.7-50-01

The Control Room Ventilation System consists of the following equipment: a single filter unit consisting of two roughing filters, two high efficiency particulate air (HEPA) filters; two activated charcoal adsorbers for removal of gaseous activity (principally iodines); two 100% capacity filter booster fans; and, a single duct system including dampers, controls and associated accessories to provide for three different air flow configurations. The air-conditioning units associated with the CRVS are governed by LCO 3.7.12, "Control Room Air Conditioning System (CRACS)."

The CRVS is divided into two trains with each train consisting of a filter booster fan and the associated inlet damper and the following components which are common to both trains: the control room filter unit, damper A (filter unit bypass for outside air makeup to the Control Room), damper B (filter unit inlet for outside air makeup to the Control Room), damper C (filter unit inlet for reticulated air), and the toilet and locker room exhaust fan. The two filter booster fans (F 31 and F 32) are powered from safeguards power trains 5A (EDG 33) and 6A (EDG 32), respectively. E&Kc4the automatic dampers that are common to both trains *positioned in the fail-safe position (open or closed) by eithe of the redundant actuation channels.

l 6


INSERT: B 3.7-50-02

The three different CRVS air flow configurations are as follows:

a) Normal operation consists of approximately 85% (8500 cfm) unfiltered recirculated flow driven by the air-conditioning fans and approximately 15% (1500 cfm) unfiltered outside air makeup;

b) Incident mode with outside air makeup (i.e. 10% incident mode) consists of approximately 87% (9250 cfm) unfiltered recirculated flow driven by the two safety related air-conditioning fans, at least 10% (> 1000 cfm) filtered recirculated flow driven by either one of the two filter booster fans and approximately 2.5% to 4.0% (250 to 400 cfm) filtered outside air makeup;

c) Incident mode with no outside air makeup (i.e. 100% incident mode) consists of 85% (9100 cfm) unfiltered reticulated flow driven by the two safety related air-conditioning fans, approximately 15% filtered recirculated flow driven by either one of the two filter booster fans and no outside air makeup.

Note that the required recirculation rates are demonstrated with .surveillance tests conducted with the air conditioning system (CRACS) operating. An inoperable CRACS fan will affect the flow balance of the CRVS due to interconnected ductwork. Therefore, if the fan associated with one of the air-conditioning units governed by LCO 3.7.12 is inoperable, Conditions in both LCO 3.7.11, Control Room Ventilation System, and LCO 3.7.12, Control Room Air Conditioning System (CRACS), will apply.

Incident mode with outside air makeup is the preferred method of operation during any radiological event because it provides outside air for pressurization of the Control Room. Calculations indicate that very low volumes of outside air makeup will maintain the Control Room at a slight positive pressure. Nevertheless, due to the difficulty of adjusting and maintaining the flow dampers to provide a low flow, it was determined that the damper should be adjusted to provide a flow of approximately 250 cfm (2.5% outside air makeup). However, a higher volume of outside air makeup to the Control Room increase the thyroid dose to the operators during an accident. Therefore, the Control Room dose assessment assumes a filtered outside air makeup of approximately 400 cfm (4.0% outside air makeup).


INSERT: B 3.7-50-02 (continued)

On a Safety Injection signal or high radiation in the Control Room (Radiation Monitor R-1), the CRVS will actuate to the incident mode with outside air makeup (i.e. 10% incident mode). This will cause one of the two filters booster fans to start, the locker room exhaust fan to stop, and CRVS dampers to open or close as necessary to filter all incoming outside air and direct approximately 10% of the recirculated air through the filter unit. In the event that the first booster fan fails to start, the second booster fan will start after a predetermined time delay.

If for any reason it is required or desired to operate with 100% recirculated air (e.g., toxic gas condition is identified), the CRVS can be placed in the incident mode with no outside air makeup (i.e. 100% incident mode) by remote manually operated switches. The Firestat detectors will also initiate 100% incident mode in the CRVS.

8 3.7. 0

BASES

BACKGROUND (continued) Outside a• is/filIter) , * d lted/ Kth buyllding/air fr e, eelectri otl e ip n a c • 1 I readinj( rooo , a u • L . •

Pre ur io henqreci t~he L&d frnm tho mtn n Pre~uiz, l~io of~te :.trol ro pevents nfil L~atid~n of tMilt er d a r f o tod su r di n a e sof th .bui• i .(

Se L-i ns omnin/ he tut ic ga ;Isolion s ate r•eth s , e x c e p t / th a t J te i g ~ l sw i • h s p o t o / o

from enter n t• co I r o . " "/ cTh ix :-t:-n t: ontrol roomi~niuul oioe

by radiation and toxic gas detectors. (ne detectr a uol 4ýe ý ý ,ýr in-will cause actuation of the emerency _radiation state-losP ,ic o on•_~ea~d~e)"

QlW -actions of the toxic gas isolation state are more restrictive, and will override the actions of the emergency radiation state.

I ----•-A Asin le train,will plr (s rz Ar~ Ior n ry- t n •h5 u• Z-E0. -. a The L-rf operation in

maintaining the control room habitable is discussed in the L~37-I-O'4 FSAR, Section It22Q (Ref. 1.~ j

redunoant supply ad recirculatio• trains rovide e req ~red liltratin soul• an exc ssive •essure op Sd•elo •ross t e other/filter ra'n. Xormally open ?•lati ~n dampe s are iuranged In series pairs •o that/the (f all • e of on e d m r Io Ah) ( l .o r l •n h r. a h o

Thesp is designed in accordance with Seismic a I requirements.

The( is designed to maintain the control room environment for 30 days of continuous occupancy after a Design Basis Accident (DBA) without exceeding a 5 rem whole

ýb~ody do~se orý oayaý•f(cPdY

APPLICABLE The CK omponents are arran ne Vin redundant, safety SAFETY ANALYSES .,Te e1 ventilation trains. e location of components and Sducting within the control J envelope ensures an adequate

su 1- of filtered air to all areas requiring access. The _.. •provides airborne radiological protection for the

control room operators, as demonstrated by the control room . accident dose analyses for the most limiting design basis

-#o"ti

Rev 1, 04/07/95

jet1

! I

WOG STS B 3.7-51


INSERT: B 3.7-51-01

On a Safety Injection signal or high radiation in the Control Room (Radiation Monitor R-1),

INSERT: B 3.7-51-02

of the CRVS (i.e., incident mode with outside air makeup (i.e. 10% incident mode)).

INSERT: B 3.7-51-03

The CRVS does not actuate automatically in response to toxic gases. Separate chlorine, ammonia and oxygen probes are provided to detect the presence of these gases in the outside air intake. Additionally, monitors in the Control Room will detect low oxygen levels and high levels of chlorine and ammonia. The CRVS may be placed in the incident mode with no outside air makeup (i.e. 100% incident mode) to respond to these conditions. Instrumentation for toxic gas monitoring is governed by the IP3 Technical Requirements Manual (TRM) ( Ref. 4). Generally, the manually initiated

INSERT: B 3.7-51-04

create a slight positive pressure in the control room.

-CRfEFs B 3.7.)d

BASES

APPLICABLE fission product release presented

SAFETY ANALYSES in the FSAR, Chapter MU(e. . ( c o n t i n u e d ) i a t y , Z t x c g s rf l 5

.•-S.-}- 2-0Y The worst case s atctiverfailure of anoemponentno orf the

•-k( _S assuming a loss of offsite power, does not impair he

f ability of the system to perform its design function.

T;he C .e:F satisfies Criteri6n 3 of oAKPNK2= i~

LCO Two Enged n trains are required to be OPERABLE to ensure that at least one is available.

umi" 994 .... e .. ..cth thkti. Total

system failure could result in exceedina a dose of 1

tne control room operator in the event of a large ýradioactive release.

"ýThe i~s considered OPERABLE when the individual com onents necessary to limit operator exposure are OPERABLE

oth ains. A ý- _ train is OPERABLE when the associated:

a.- jj~f OPERABLE;

~b. HEPA filters and charcoal adsorbers are not

excessively restricting flow, and are capable of performing their filtration functions; and

c. •ea r d i t ru ctwork, valves, and dampers are

LOPEALE, and air circulation can be maintained.

In addition, the control room boundary must be maintaine(

,including the integrity of the walls, floors, ceilings, j• 3,7-•2o06 ductwork, and access doors.

9,

APPLICABILITY In MODES 1, 2, 3, 4,[5an 1,/ adurnlymov fint f

a mbes [aO dui ng COR Av• T•tr]

__ja-teoqe u

WOGSISB .7-2 ev , 4/07/95ud-

Rev 1, 04/07/95B 3.7-52WO STS


INSERT: B 3.7-52-01

accident (i.e., DBA LOCA)

INSERT: B 3.7-52-02

Radiation monitor R-1 is not required for the Operability of the Control Room Ventilation System because control room isolation is initiated by the safety injection signal in MODES 1, 2, 3, and 4 and control room isolation is not required for maintaining radiation exposure within General Design Criteria 19 limits following a fuel handling accident or gas-decay-tank rupture.

INSERT: B 3.7-52-03

However, the original CRVS design was not required to meet single failure criteria and, although upgraded from the original design, CRVS does not satisfy all requirements in IEEE-279 for single failure tolerance. Each of the automatic dampers that are common to both trains is positioned in the fail-safe position (open or closed) by either of the redundant actuation channels.

INSERT: B 3.7-52-04

5 rem whole body or 30 rem to the thyroid of

INSERT: B 3.7-52-05 F `-<ilter booster fan and an air-conditioning unit fan powered from the

same safeguards power train are


INSERT: B 3.7-52-06

Instrumentation for toxic gas monitoring is governed by the IP3 Technical Requirements Manual (TRM) ( Ref. 4) and is not included in the LCO.

Note that the required recirculation rates are demonstrated with surveillance tests conducted with the air conditioning system (CRACS) operating. An inoperable CRACS fan will affect the flow balance of the CRVS due to interconnected ductwork. Therefore, if the fan associated with one of the air-conditioning units governed by LCO 3.7.12 is inoperable, Conditions in both LCO 3.7.11, Control Room Ventilation System, and LCO 3.7.12, Control Room Air Conditioning System (CRACS), will apply.

B 3.7.,W

APPLICABILITY Qj6must be OPERABLE to ~ v operator exposure during

(continued) and following a DOBA.

In (-J5 or ,the oCRE $4s required cope with e re .aise from e rupture f an outsid aste gas k.

During ement of adiated fu assemblies and COR a3T-Vl ALTE seONS], the yEfFS must b PERrABLE to e wit he

rel se from a fiel handl ing ccidentr

ACTIONS A.IWhen one train is inoperable, action must be taken to

CiV U restore OPERABLE status within 7 days. In this Condition, e remaininglOPERABLE (CR.S train is adequate to perform

the control room protection function. However, the overall reliabilit is reduced be e a -"wy4e failure in the

Ltrain could resu t in oss o RE function. The 7 day Completion Time is based on the low pro ability of a DBA occurring during this time period, and ability of the

Sremaining train to provide the required capability.

Te�t

83.7-S�oz � �1 an�2 �

in I, 2,,2 ,theo no-era REFS n cajomt) UlkfeAJtoe. -sOPE aB1stiai within the required Completion Time, the unit must be placed in a MODE that

7 31 yb minimizes accident risk. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

C andt.z

MODE 5 r 6, or] r4ing moveme of irradled fuel ssembli [, or during CORE AL RATIONS], the inopevble

CREFS ain canno'be restore to OPERABLFstatus witfn the req•" rerd Compl-t"on Time, aation must b taken to 7l iumiediately place the OPERABLE CREFS tain in the emergency

(continued)-

Rev 1, 04/07/95

7

WOG STS B 3.7-53


INSERT: B 3.7-53-01

The CRVS is not required in MODE 5 or 6, or during movement of irradiated fuel assemblies and core alterations because analysis indicates that isolation of the control room is not required for maintaining radiation exposure within acceptable limits following a fuel handling accident or gas decay tank rupture.

Administrative controls address the role of the CRVS in maintaining control room habitability following an event at Indian Point Unit 2.

INSERT: B 3.7-53-02

B.1

When neither CRVS train is Operable, action must be taken to restore at least one train to OPERABLE status within 72 hours. The 72 hour Completion Time is acceptable because of the low probability of a DBA occurring during this time period.

INSERT: B 3.7-53-03

If Required Actions A.1 or B.1 are not met

-eRf1sB 3.7. JO

BASES

ACTIONS an - (co tinted)mode. This •tion ens. es that the r inn traip/is. OPERBLE, •at no fai res preventi.rAv automatic 43tuation Will oc , and tha~anay active f lure would be readily

An lternative o Required Ation C.1 is U6 ?uiedia y si ties that c•d1d result ini& release f

radioactiv y that mighVrequire isol$ion of t control room. TI s places th nit in a condition tha minimizes

risk. is does notj reclude the Jovement oefuel to a safe posit .n. I //

R uired Actio C.1 is modif ed by a N t(e indicating to

ace the sy em in the to c gas prpt9ection mode if

automatic ansfer to to c gas protection mode is

inoperabl.

'n MODE 5 or , or] dur' g movement of radiated fuel assembliesY, or durzn CORE ALTERATIO 3,with two CRE trains interable, arion nust be tin imediately2 t3 suspend/ictiv i ties at could resur in a reiease o radioa~tivity thai/might enter tl control room. This

)lais the unit i'n a conditionthat minimizes ac-6ident risk.

Fs does not p~reclude the movement of fuel tcr'a safe .

pos it ion.__ __:

If bothA•EFS t~tins are/4noperab nv'n MODE .K2, 3, or¥,

he EF$maydotbe able of, erforming •e intendr \

ucion a ~the un•" is in a ondition tside the- ccidentt

l~ses /Theref 'e, LC0 3 .3 must b entered i diately.J

SURVEILLANCE SR 3.7.Z '.I REQUIREMENTS

Standby systems should be checked periodically to ensure

that they function properly. As the environment and normal

operating conditions on this system are not too severe,

..- M

Rev 1, 04/07/95B 3.7-54WOG STS

ewif 8 3.7.W

BASES I'

SURVEILLANCE SR 3.7.1e.1 (continued) REQUIREMENTS testing each train once ever month rovides an adequate

check of this asetem._ rla t ofhe h moisture w tainurdundat n theav alat Thisnt . [SS verems t iha heatere r fsorui

pe rf ntin uouhours w ith the heatia energil Ft Tstin

wiout heotrs need 1y be opera d for inut d

monstratgthe p once o the s.fi&--lter 31 day

Frequency is based on the reliability ol t equipment and

the two train redundancy availability.

II

SR 3.7.I1.2

SR This SR verifies that the requiredai testing os

performed in accordance with the u Ventilation silter F esting

Thiso SR veriFies The fintegritests ore intahecctro ce wi •Re-gulator• Guide 1.52 (Ref. The ,[VFTPJ ir •nclu

esting the performance of the HEaA filter, charcoal

pasorber efficiency, minimum flow rate, and tr e poysicav

properties of the activated charcoal. Specific test a Frequencies and additional information are discussed in

detail in the k(VFTPý.

SR 3.7. 5.3 /•This SR verifies that each(týStrain starts and operates

on an actual or simulated actuation signal. The Frequency

of rMmonths is e

SR 3.7.1-0.4

This SR verifies the integrity of the control room

enclosure, and the assumed inleakage rates of the

potentially contaminated air. The control room positive

pressure, with respect to potentially contaminated adjacent

area s priodically tested to verify proper functioning -- f-the P) During the~mLg'. _ ' at the

V' i d--O. s-igned tol res ! t,4postive pressure with respect

t-o a'a~n area in rder't prevent unfiltered inleakage. S•)F Ther is designed torrmainntain this positive pressure

Rev 1, 04/07/95B 3.7-55WOG STS


INSERT: B 3.7-55-01

based on operating experience which has demonstrated this Frequency provides a high degree of assurance that the booster fans will operate and dampers actuate to the correct position when required.

INSERT: B 3.7-55-02

operation in the incident mode with outside air makeup (i.e. 10% incident mode)

CREFS B 3.7.10

BASES

SURVEILLANCE SR 3.7. W-4 (continued) REQUIREMENTSIkep f li- ate 3 Th e

1:0•. ", -•Fre ency of [I• mont,) on a -TAGGED TE TBASIC s TW. csis nt -ji the goýdance Zrovid nNFG2

(it2ý ef.)

REFERENCES 1.

2.

3.

FSAR, Section

FSAR, Chapter

Regulatory Guide 1.52, Rev. 2.

B 3.7-56 Rev 1, 04/07/95WOG STS


INSERT: B 3.7-56-01

with very low volumes of outside air makeup. Due to the difficulty of adjusting and maintaining the flow dampers to provide a low flow, it was determined that the damper should be adjusted to provide a flow of approximately 250 cfm (2.5% outside air makeup). Note that the higher the volume of outside air makeup to the Control Room, the higher the thyroid dose to the operators during an accident. The acceptance criteria of 400 cfm (4.0% outside air makeup) is the volume used in the Control Room dose assessment.

The SR Frequency of 24 months on a staggered test basis is acceptable because operating experience has demonstrated that the control room boundary is not normally disturbed. Staggered testing is acceptable because the SR is primarily a verification of Control Room integrity because fan operation is tested elsewhere.

INSERT: B 3.7-56-02

4. IP3 Technical Requirements Manual


Conversion Package

Technical Specification 3.7.12: "Control Room Air Conditioning System (CRACS)"

PART 1:



a O-V I a Kt, cf.( 11 -j4-r ? 0, De- C)n I I

CRACS B 3.7.12

BASES

ACTIONS (continued)

When neither CRACS train is Operable, action must be taken to restore at least one train to OPERABLE status within 72 hours. The 72 hour Completion Time is acceptable because of the low probability of a DBA occurring during this time period and because alternate nonsafety cooling means are typically available.

C.1 and C,2

If Required Actions A.1 or B.1 are not met within the required Completion Time, the unit must be placed in a MODE that minimizes the risk. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.


This SR verifies that the heat removal capability of the system is sufficient to remove the heat load required to maintain functional capacity of the Control Room at all times (Ref. 1). This SR consists of a combination of testing and calculations. The 24 month Frequency is appropriate since significant degradation of the CRACS is slow and is not expected over this time period.




Conversion Package

Technical Specification 3.7.12: "Control Room Air Conditioning System (CRACS)"

PART 5:


NUREG-1431 and ITS


OG No. TSTF No. Generic Change Description NRC STATUS IP3 STATUS JD No.

BWROG-017 051 RO REVISE CONTAINMENT NRC Review Not Incorporated NIA REQUIREMENTS DURING HANDLING IRRADIATED FUEL AND CORE ALTERATIONS (REQUIREMENTS LIMITED TO "RECENTLY" IRRADIATED FUEL)


-GRE3.. 8 3.7. 4

BASES

ACTIONS (continued)


SR 3.7,V.1

This SR verifies that the heat removal capability of the system is sufficient to-remove the heat load assumed in thc E[6fety flysesj] in the c-ntroal- m. This SR consists of a combination of testing and calculations. The month Fre:qu is appropriate since significant degradation oF-the is slow and is not expected over this time' period.

REFERENCES 1. FSAR, Section

Rev 1, 04/07/95

!

WOG STS B 3.7-60

NUREG-1431 Markup Inserts ITS SECTION 3.7.12 - Control Room Air Conditioning System (CRACS)

INSERT: B 3.7-60-01

required to maintain functional capacity of the Control Room at all times (Ref. 1). IftJ


Conversion Package

Technical Specification 3.7.15: "Spent Fuel Pit Boron Concentration"

PART 1:



Spent Fuel Pit Boron Concentration 3.7.15



SR 3.7.15.1 Verify the spent fuel pit boron concentration 31 days is within limit.

INDIAN POINT 3 Amendment [Rev.1], 06/29/00

I,(-

3.7.15-2

Spent Fuel Pit Boron Concentration B 3.7.15

BASES


The concentration of dissolved boron in the spent fuel pit satisfies Criterion 2 of 10 CFR 50.36.

LCO The spent fuel pit boron concentration is required to be > 1000 ppm. The specified concentration of dissolved boron in the spent fuel pit preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference 3. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the spent fuel pit until a spent fuel pit verification confirms that there are no mis-loaded fuel assemblies. With no mis-loaded fuel assemblies and unborated water, the spent fuel pit design is sufficient to maintain the core at kff < 0.95.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pit, until a complete spent fuel pit verification has been performed following the last movement of fuel assemblies in the spent fuel pit. This LCO does not apply following the verification, since the verification would confirm that there are no misloaded fuel assemblies. With no further fuel assembly movements in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly.

I

A.1. A.2.1 and A,2.2

The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentration of boron in the spent fuel pit is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending

(continued)

INDIAN POINT 3 Revision [Rev.1], 06/30/00

ACTIONS

B 3.7.15 -3

Spent Fuel Pit Boron Concentration B 3.7.15

BASES

ACTIONS A.1. A.2.1 and A.2.2

movement of fuel assemblies. Alternatively, beginning a verification of the Spent Fuel Pit fuel locations, to ensure proper locations of the fuel, can be performed. However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.

If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2. 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.


SR 3.715.1

This SR verifies that the concentration of boron in the spent fuel pit is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The 31 day Frequency is appropriate because no major replenishment of spent -I

fuel pit water is expected to take place over such a short period of time. This SR is not required to be met or performed if a spent fuel pit verification for conformance with LCO 3.7.16; Figures 3.7.16-1 and B 3.7.16-1. has been performed on all fuel assemblies since the last verification following the last movement of fuel assemblies in the spent fuel pit.

REFERENCES 1. Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

(continued)



Conversion Package

Technical Specification 3.7.15: "SPENT FUEL PIT BORON CONCENTRATION"

PART 2:

CURRENT TECHNICAL SPECIFICATION PAGES

Annotated to show differences between CTS and ITS

CTS PAGE 3.8-3 3.8-5 3.8-6

AMENDMENT FOR REV 0 SUBMITTAL

114 173 175

AMENDMENT FOR REV 1 SUBMITTAL

114 173 189

3.8-7 173 189

T4.1-2(1) 139 200

5.4-1 173 173

COMMENT

Deleted Requirement of 267 Hours for Discharge of > 76 Assemblies.

No impact on 3.7.15. Deleted FSAR Reference in Bases Page. No impact on 3.7.15.

Deleted Boric Acid Tank Sampling Requirement Frequencies

No impact on 3.7.15


TARIM A 1 .9 - kl -1l £ UL )

1. Reactor Coolant

2. Boric Acid Tank

3. Spray Additive Tank

4. Acecmulators

5. Refueling vater Storage Tank

6 . Secondary Coolant

7. Cop o--nt Cooling Vtert

I � I

_ ?P tFuel Pool

F~mEnVul" FOR SAM? -NO TSU TU

Cross Activity(• Tritium Activity Boron concentration tadiochemlcal (Sam,)"'2

Spectral Check Oxygen and Chlorides

Concentration Fluorides Concentration

r Ceterncnation (81 Isotopic Analysis for 1-131,. - 3 , -3

Boron Concentration. Chlorides HoNI Conceontration

Boron Cone•.ntration

Boron Coneontration PHg. Chlorides

Gross Activity

1-131 Equivalent (liotopic Analysis)

Gross Activity

Cross Activity. Corrosion Inhibitor and pH

i~~ LI2!L~~vty Borona .. 9 n s I f---'-"- I

Freauency

5 days/week("1 4 41 Weekly"') 2 days/veek Monthly

3 times per 7 days

VWckly

Semi -Annually Once per 14 dayell)

eekhly

monthly

Monthly

monthly

Quarterly I

IMonthly

3 time. par 7 day.

Monthly

3 day."'1 10 days 5 days

45 days

3 days

10 days

30 weeks 20 days

lO days

10 day.43 days

45 days

45 days

45 days

3 days

i 4.5 day.-... 5

Amendment Ho 139 \b 04'r 7Sf

XT

-

C1T5

HORKUP

8k $3.7-16. (

CA)

V-- A

i

?AIIIJIE • I.• lek.-• : -m 4.

n|AJnnl•tla •* . ...........

116 weeksQuarterly


Conversion Package


PART 3:




DISCUSSION OF CHANGES ITS SECTION 3.7.15 - SPENT FUEL PIT BORON CONCENTRATION

MORE RESTRICTIVE

None

LESS RESTRICTIVE

L.1 CTS 3.8.C.3 specifies the Applicability of requirements for spent fuel pit boron concentration as during periods of spent fuel movement in the spent fuel pit. CTS 5.4.3 specifies the Applicability of requirements for spent fuel pit boron concentration as whenever there is fuel in the pit (See ITS 3.7.15, DOC A.3). CTS 4.1-2, Item 8, requires monthly verification of boron concentration "when fuel stored" in the spent fuel pit. Therefore, IP3 requirements for spent fuel pit boron concentration apply whenever there is fuel in the spent fuel pit.

ITS 3.7.15 specifies that spent fuel pit boron concentration limits must be met "When fuel assemblies are stored in the fuel storage pit and a spent fuel pit verification has not been performed since the last movement of fuel assemblies in the fuel storage pit." This change is -acceptable because, as stated in the SER Related to IP3 CTS Amendment 173, April 15, 1997, the spent fuel pit design is sufficient to maintain the core at keff • 0.95 with unborated water if there are no misloaded fuel assemblies (i.e., fuel storage locations meet the requirements of ITS LCO 3.7.16). The requirement in ITS SR 3.7.16.1 to verify proper storage location of each fuel assembly prior to landing the assembly in the spent fuel pit and the requirement to perform a complete spent fuel pit verification prior to relaxing boron concentration requirements provides a high degree of assurance that boron concentration will be maintained at all times there is a potential for a mis-positioned fuel assembly. Finally, even if requirements for boron concentration are eliminated, spent fuel pit boron concentration is not reduced after the completion of fuel movement and is usually maintained > 1900 ppm. As a result, the practical consequence of this change is to eliminate the formal requirement in CTS Table 4.1-2, Item 8, for monthly verification of spent fuel pit boron concentration after a storage location verification is completed. Therefore, eliminating the requirements for a minimum boron concentration in the spent fuel pit if no fuel movement

Indian Point 3 1 TTZ -.. -4.,a_ 1 rI..W UlIIV%=IlUr 3UUIIIILLdl, Kev I•J

DISCUSSION OF CHANGES ITS SECTION 3.7.15 - SPENT FUEL PIT BORON CONCENTRATION

is in progress and a storage verification was performed after the last movement of fuel assemblies will not significantly increase the potential for criticality in the spent fuel pit. This change has no significant adverse impact on safety.

REMOVED DETAIL

None

Indian Point 3 ITS Conversion Submittal, Rev 14


Conversion Package


PART 4:





NO SIGNIFICANT HAZARDS EVALUATION ITS SECTION 3.7.15 - Spent Fuel Pit BORON CONCENTRATION

LESS RESTRICTIVE ("L.1" Labeled Comments/Discussions)

New York Power Authority has evaluated the proposed Technical Specification change identified as "Less Restrictive" in accordance with the criteria set forth in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The bases for the determination that the proposed change does not involve a significant hazards consideration are discussed below.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

This change eliminates the requirement for maintaining a minimum boron concentration in the spent fuel pit if no fuel movement is in progress and a storage verification was performed after the last movement of fuel assemblies. This change will not result in a significant increase in the probability or consequences of an accident previously evaluated because , as stated in the SER Related to IP3 CTS Amendment 173, April 15, 1997, the spent fuel pit design is sufficient to maintain the core at keff • 0.95 with unborated water if there are no misloaded fuel assemblies (i.e., fuel storage locations meet the requirements of ITS LCO 3.7.16). The requirement in ITS SR 3.7.16.1 to verify proper. storage location of each fuel assembly prior to landing the assembly in the spent fuel pit and the requirement to perform a complete spent fuel pit verification prior to relaxing boron concentration requirements provides a high degree of assurance that boron concentration will be maintained at all times there is a potential for a mis-positioned fuel assembly. Finally, even if requirements for boron concentration are eliminated, spent fuel pit boron concentration is not reduced after the completion of fuel movement and is usually maintained > 1900 ppm. As a result, the practical consequence of this change is to eliminate the formal requirement in CTS Table 4.1-2, Item 8, for monthly verification of spent fuel pit boron concentration after a storage location verification is completed. Therefore, eliminating the requirements for a minimum boron concentration in the spent fuel pit if no fuel movement is in progress and a storage verification was performed after the last movement of fuel assemblies will not significantly increase the potential for criticality in the spent fuel pit.

Indian Point 3 1 TTS rnnvarcinn Q ,hmif+ I D',, 1± •. •UII •-I•JI I U II M U, i ,( I z • V IL

NO SIGNIFICANT HAZARDS EVALUATION ITS SECTION 3.7.15 - Spent Fuel Pit BORON CONCENTRATION

2. Does the change create the Dossibility of a new or different kind of accident from any accident previously evaluated?

The proposed change will not involve any physical changes to systems, structures, or components, or involve a change in normal plant operation. Therefore, it will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

This change does not involve a significant reduction in a margin of safety because, as stated in the SER Related to IP3 CTS Amendment 173, April 15, 1997, the spent fuel pit design is sufficient to maintain the core at keff • 0.95 with unborated water if there are no misloaded fuel assemblies (i.e., fuel storage locations meet the requirements of ITS LCO 3.7.16). The requirement in ITS SR 3.7.16.1 to verify proper storage location of each fuel assembly prior to landing the assembly in the spent fuel pit and the requirement to perform a complete spent fuel pit verification prior to relaxing boron concentration requirements provides a high degree of assurance that boron concentration will be maintained at all times there is a potential for a mis-positioned fuel assembly. Finally, even if requirements for boron concentration are eliminated, spent fuel pit boron concentration is not reduced after the completion of fuel movement and is usually maintained at 1900 ppm. As a result, the practical consequence of this change is to eliminate the formal requirement in CTS Table 4.1-2, Item 8, for monthly verification of spent fuel pit boron concentration after a storage location verification is completed. Therefore, eliminating the requirements for a minimum boron concentration in the spent fuel pit if no fuel movement is in progress and a storage verification was performed after the last movement of fuel assemblies will not significantly increase the potential for criticality in the spent fuel pit.



Conversion Package


PART 5:


NUREG-1431 and ITS


OG No. TSTF No.

CEOG-023 R1 070 R1

Generic Change Description

FUEL STORAGE POOL VERIFICATION

NRC STATUS

APPROVED/INCO RPORATED

CEOG-051 139 RI INCORRECT CRITERIA DEFINED APPROVED/INCO Incorporated T.2 IN B 3.7.16 RPORATED


IP3 STATUS

Incorporated

JD No.

T.1

)

Fuel Storage Pool Boron Concentration 3.7./,

/



SR 3.7. W.1 Verify the isuE3325Witilnboron ldats /5'• concentration is within limit. /*

-J,4;p- 1 1)

I Q. 1

Rev 1, 04/07/95WOG STS 3.7-37

Fuel Storage Pool Boron Concentration

B 3.7. Z

BASES

APPLICABILITY (continued)

ACTIONS

progress, there is no potentialfor a misloaded fuel assembly or a dropped fuel assembly.

A.I. A.2.1. and A.2.2

The Required Actions are modified by a Note indicating that

LCO 3.0.3 does not apply.

When the concentration of boron in the fuel storage pool is

less than required, immediate action must be taken to

preclude the occurrence of an accident or to mitigate the

consequences of an accident in progress. This is most

efficiently achieved by immediately suspending the movement

of fuel assemblies. The concentration of boron is restored

simultaneously with suspending movement of fuel assemblies.

An pceptaoi altern tYU o verl Dy aumiistrat e •pns thaJh flstrgpool vp-1ficatlhsben

tperfo•' sinc•_ e .lasVrimvemento°ffulsshlsinh L~uel~traae •!~l However, prior to restumngi vemO ent, of

fuel assemblies, the concentration of boron must be

restored. This does not preclude movement of a fuel

assembly to a safe position.

If the LCO is not met while moving irradiated fuel

assemblies in MODE 5 or 6, LCO 3.0.3 would not be

applicable. If moving irradiated fuel assemblies while in

MODE 1, 2, 3, or 4, the fuel movement is independent of

reactor operation. Therefore, inability to suspend movement

of fuel assemblies is not sufficient reason to require a

reactor shutdown.

SURVEI LLANCE S .S-aREQUIREMENTS

This SR verifies that the concentration of boron in the fuel

storage pool is within the required limit. As long as this

SR is met, the analyzed accidents are fully addressed. The

(,.?'day Frequency is appropriate because no major

replenishment of pool water is expected to take place over

such a short period of time.

B 3.7-83

(continUea)(contRnue4 )

Rev 1, 04/07/95

'K,' <�CL3I f>

WODG STS

REQUIREHENTS

NUREG-1431 Markup Inserts ITS SECTION 3.7.15 - SPENT FUEL PIT BORON CONCENTRATION

INSERT: B 3.7-83-01

Alternatively, beginning a verification of the Spent Fuel Pit fuel locations, to ensure proper locations of the fuel, can be performed.

INSERT: B 3.7-83-02

This SR is not required to be met or performed if a spent fuel pit verification for conformance with LCO 3.7.16, Figures 3.7.16-1 and B 3.7.16-1, has been performed on all fuel assemblies since the last verification following the last movement of fuel assemblies in the spent fuel pit.

I


Conversion Package


PART 6:

Justification of Differences between

NUREG-1431 and IP3 ITS


JUSTIFICATION OF DIFFERENCES FROM NUREG-1431 ITS SECTION 3.7.15 - SPENT FUEL PIT BORON CONCENTRATION

RETENTION OF EXISTING REOUIREMENT (CURRENT LICENSING BASIS)

CLB.1 CTS Table 4.1-2, Item 8 requires a monthly verification of spent'fuel pit boron concentration whenever fuel is stored in the spent fuel pit. NUREG-1431, STS SR 3.7.16 requires verification of spent fuel pit boron concentration every 7 days unless a complete fuel location verification is completed. ITS SR 3.7.15 will maintain the current license basis of performing this surveillance every 31 days. NYPA would normally option to maintain the minimum boron concentration all the time because it is an onerous task to perform the complete verification of all the spent fuel assemblies due to the numerous spent fuel assemblies. Maintaining the current license basis of performing this surveillance every 31 days is needed because a 7-day Frequency is a burden without an appreciable improvement to safety. Normally the spent fuel pit is maintained at > 1900 PPM of boron concentration that is well above the minimum required before movement of irradiated fuel assemblies is allowed by ITS 3.7.15 (i.e., > 1000 PPM). There is a high degree of assurance that the spent fuel pit minimum boron cocnentration will be maintained above the LCO limits because it is very unlikely that a significant dilution of the borated water in the spent fuel pit would take place over this period of time. This change has no significant adverse impact on safety.

PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT

PA.1 Corrected typographical error or made a minor editorial improvement to improve clarity and ensure requirements are fully understood and consistently applied. There are no technical changes to requirements as specified in NUREG 1431, Rev. 1; therefore, this change is not a significant or generic deviation from NUREG 1431, Rev 1.

PLANT-SPECIFIC DIFFERENCE IN THE DESIGN OR DESIGN BASIS

DB.1 Design or implementation details are incorporated or revised as necessary to more precisely describe IP3 current design or practice. These changes are intended to describe the design, improve clarity, or ensure requirements are fully understood and consistently applied. Unless identified and described below, these changes are self-



Conversion Package

Technical Specification 3.7.16: "Spent Fuel Assembly Storage"

PART 5:


NUREG-1431 and ITS


OG No. TSTF No. Generic Change Description NRC STATUS IP3 STATUS JD No. N/A N/A NO GENERIC CHANGES ARE Not Applicable Not Applicable N/A

POSTED AGAINST THIS SPECIFICATION.


NUREG-1431 Markup Inserts ITS SECTION 3.7.16 - SPENT FUEL ASSEMBLY STORAGE

INSERT: B 3.7-85-02

and fuel storage locations, enrichment and burnup are in conformance with analysis assumptions and this LCO.

INSERT: B 3.7-85-03

concentration is verified to be within the limits specified in LCO 3.7.15, Spent Fuel Pit Boron Concentration, prior to movement of f4a-assembl es in the see-t atiel pi.

INSERT: B 3.7-85-04

(i.e., verification that the spent fuel pit boron concentration is within limit).

(i,. '


Conversion Package

Technical Specification 3.7.17: "Secondary Specific Activity"

PART 3:



Indian Point 3 ITS Submittal, Revision 1 . - Il

DISCUSSION OF CHANGES ITS SECTION 3.7.17 - SECONDARY SPECIFIC ACTIVITY

limit in Modes 1, 2, 3 and 4. In conjunctionwith the change in applicability, ITS 3.7.17, Required Action A.1 and A.2 require that the unit be placed in and Mode 5 (i.e., outside the expanded Applicability) if the specific activity is not within the limit.

This change, adding Mode 4 to the Applicability for secondary system specific activity, is needed because SG venting to the atmosphere may be needed to remove decay heat in Mode 4. This change is acceptable because it does not introduce any operation which is un-analyzed while requiring that the LCO for secondary system specific activity is applicable whenever there is a potential that SG venting to the atmosphere may be needed to remove decay heat. Therefore, this change has no adverse impact on safety.

M.2 CTS Table 4.1-2, Item 6, requires verification that secondary system specific activity (i.e., Dose Equivalent 1-131) is within specified limits every month with a maximum time between secondary coolant specific activity analyses of 45 days. ITS 3.7.17 maintains the requirement to verify secondary system specific activity every 31 days but limits the maximum time between analyses based on ITS SR 3.0.2 which allows a 25% grace period for a maximum interval of approximately 7.5 days. This change is not needed to satisfy technical requirements but is being adopted for consistency with the NUREG-1431 and to simplify application of ITS SR 3.0.2. This change has no impact on safety.

LESS RESTRICTIVE

L.1 CTS Table 4.1-2, Item 6 (second part), requires checking secondary system gross activity at least 3 times every 7 days with a maximum time between secondary gross activity analyses of 3 days. ITS 3.7.17, Secondary Specific Activity, does not have any requirements for periodic verification of secondary side gross activity. Deletion of the requirement for periodic verification of secondary side gross activity is acceptable because limitations on SG gross activity is not an assumption of any accident analysis and the information provided by this test (i.e., early identification of SG tube leakage and significant

2 ITS Conversion Submittal, Rev 1Indian Point 3

DISCUSSION OF CHANGES ITS SECTION 3.7.17 - SECONDARY SPECIFIC ACTIVITY

changes in RCS activity) is provided to theIoperators by other methods (e.g., air ejector monitors). Therefore, deletion of this requirement has no adverse impact on safety.

REMOVED DETAIL

None



Conversion Package


"Secondary Specific Activity"

PART 4:





NO SIGNIFICANT HAZARDS EVALUATION ITS SECTION 3.7.17 - SECONDARY SPECIFIC ACTIVITY

LESS RESTRICTIVE ("L.1" Labeled Comments/Discussions) York Power Authority has evaluated the proposed Technical Specification change identified as "Less Restrictive" in accordance with the criteria set forth in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The bases for the determination that the proposed change does not involve a significant hazards consideration are discussed below. 1. Does the change involve a significant increase in the probability or

consequences of an accident previously evaluated?

CTS Table 4.1-2, Item 6 (second part), requires checking secondary system gross activity at least 3 times every 7 days with a maximum time between secondary gross activity analyses of 3 days. ITS 3.7.17, Secondary Specific Activity, does not have any requirements for periodic verification of secondary side gross activity. This change will not result in a significant increase in the probability or consequences of an accident previously evaluated because limitations on SG gross activity is not an assumption of any accident analysis and the information provided by this test (i.e., early identification of SG tube leakage and significant changes in secondary side activity) is provided to the operators by other methods (e.g., air ejector monitors).

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change will not involve any physical changes to systems, structures, or components, or involve a change in normal plant operation. Therefore, it will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

This change does not involve a significant reduction in a margin of safety because secondary side gross activity is not an assumption in any accident analysis and the information provided by sampling (i.e., early identification of SG tube leakage and significant changes in secondary side activity) is provided to the operators by other methods (e.g., air ejector monitors).



Conversion Package

Technical Specification 3.9.2: "Nuclear Instrumentation"

PART 3:




DISCUSSION OF CHANGES ITS SECTION 3.9.2 - NUCLEAR INSTRUMENTATION

ADMINISTRATIVE

A.1 In the conversion of the Indian Point Unit 3 Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS) certain wording preferences or conventions are adopted which do not result in technical changes (either actual or interpretational). Additionally, editorial changes, reformatting, and revised numbering are adopted to make ITS consistent with the conventions in NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Rev. 1, i.e., the improved Standard Technical Specifications.

The CTS Bases are deleted and replaced with comprehensive ITS Bases designed to support interpretation and implementation of the associated Technical Specifications. The Bases explain, clarify, and document the reasons (i.e., bases) for the associated Technical Specifications, and reflect the IP3 plant specific design, analyses, and licensing basis. In accordance with 10 CFR 50.36(a), the ITS Bases are included with the proposed ITS conversion application: however, deletion of the CTS Bases and the adoption of the ITS Bases is an administrative change with no impact on safety.

A.2 CTS Limiting Conditions for Operation (LCOs) and Surveillance Requirements (SRs) include statements of the objective and the applicability. The CTS statements of objective and applicability are deleted because these statements do not establish any requirements and do not provide any guidance for the application of CTS requirements. Therefore, deletion of these statements has no significant adverse impact on safety.

A.3 CTS 3.8.A.4 requires that the core subcritical neutron flux be continuously monitored by two source range neutron monitors whenever core geometry is being changed. When core geometry is not being changed, at least one source range neutron flux monitor shall be in service. CTS 3.8.B requires that no operations that may increase the reactivity of the core shall be made if CTS 3.8.A.4 is not met. ITS LCO



3.9.2 requires that two source range neutron flux monitors be operable in Mode 6. ITS LCO 3.9.2, Required Actions A.1 and A.2, require suspending Core Alterations and positive reactivity additions if only one SRM is Operable. Additionally, ITS LCO 3.9.2, Required Action B.1, requires immediate initiation of action to restore one SRM whenever both are inoperable. These are administrative changes with no impact on safety because . ITS LCO 3.9.2, Required Actions A.1, A.2 and B.1, are consistent with a reasonable interpretation of the existing requirements.

A.4 CTS 3.8.A specifies that requirements for source range monitors are applicable during handling operations and during reactor vessel head removal or installation: however, CTS 3.8.A.4 specifies that at least one source range neutron flux monitor shall be in service when core geometry is not being changed, ITS LCO 3.9.2, LCO and Applicability, requires 2 SRMs Operable in Mode 6: however, ITS LCO 3.9.2, Required Actions A.1 and A.2, require suspending Core Alterations and positive reactivity additions if only one SRM is Operable. These are administrative changes with no impact on safety because the combination of the ITS LCO 3.9.2 Applicability and Required Actions A.1 are consistent with a reasonable interpretation of the existing requirements.

MORE RESTRICTIVE

M.1 CTS 3.8.A.4 and CTS 3.8.B implicitly require suspending Core Alterations and positive reactivity additions if only one SRM is Operable and immediate initiation of action to restore at least one SRM whenever both are inopePable. However, CTS 3.8.A.4 and CTS 3.8.B do not require more frequent verification of the boron concentration of all filled portions of the RCS and the refueling canal if no SRMs are Operable. ITS LCO 3.9.2, Required Action B.2, is added to require performing SR 3.9.1.1 (i.e., verification of refueling boron concentration) once per 12 hours if two required source range neutron flux monitors are inoperable. This change is needed because SRMs, including the associated audible alarms, are assumed to be Operable for the detection of a boon dilution event in



sufficient time for operator actions needed-to-prevent an unplanned criticality..'More frequent performance of SR 3.9.2.1 has no adverse impact on safety.

M.2 CTS 3.8.A.4 requires that the subcritical core be continuously monitored by two SRMs whenever core geometry is being changed and by at least one SRM at other times in Mode 6; however, there are no surveillance requirements for the periodic calibration of SRMs during refueling. ITS SR 3.9.2.2 is added to require performance of an SRM channel calibration every 24 months. The 24 month Frequency is the same as the requirement for calibration of the SRM trip function which is required to be Operable during reactor plant startups. This change is needed because it ensures that the SRM detector plateau or pre-amp discriminator curves are consistent with manufacturers' data for SRM performance. This process ensures the Operability of the source range neutron flux monitors. This change is acceptable because it does not introduce any operation which is un-analyzed while requiring periodic verification of SRM Operability at the same frequency as when the instrument is used during plant startups. Therefore, this change has no adverse impact on safety.

M.3 CTS 3.8.A.4 specifies that only one source range monitor is required if core geometry is not being changed; therefore, only one SRM is required during a positive reactivity addition that does not involve a change to core geometry. ITS LCO 3.9.2 requires 2 SRMs Operable in Mode 6 and ITS LCO 3.9.2, Required Actions A.1 and A.2, require suspending Core Alterations and positive reactivity additions if only one SRM is Operable. Therefore, ITS LCO 3.9.2 requires 2 SRMs during a positive reactivity addition that does not involve a change to core geometry. This change is needed because it requires redundant reactivity change monitoring capability during positive reactivity additions when the reactor vessel head is removed. This change is acceptable because it does not introduce any operation which is un-analyzed while redundant reactivity change monitoring capability during positive reactivity additions when the reactor vessel head is removed. Therefore, this change has no adverse impact on safety.



LESS RESTRICTIVE

None

REMOVED DETAIL

LA.1 CTS 3.8.A.4 requires that the subcritical core be continuously monitored by two SRMs and that each of these SRMs have continuous visual indication in the control room whenever core geometry is being changed. ITS LCO 3.9.2 maintains the requirement for two SRMs in Mode 6 and that Core Alterations and positive reactivity additions must be terminated if only one SRM is Operable; however, the detail related to visual indication is not included in ITS LCO 3.9.2 and is relocated to the FSAR and ITS Bases. This change is acceptable because ITS LCO 3.9.2 still requires that SRMs are Operable and the requirements for Operability are defined in the ITS LCO 3.9.2 Bases.

This change is acceptable because ITS 3.9.2 maintains the existing requirement for the Operability of two SRM channels; therefore, there is no change to the existing requirements and no change to the level of safety of facility operation.

This change, which allows the description of the design of the SRMs to be maintained in the FSAR and the detailed description of the requirements for Operability of the SRMs to be maintained in the ITS Bases, is consistent with the approach used in NUREG-1431 for all Limiting Conditions for Operation (LCOs). This approach is acceptable because the requirements of 10 CFR 50.59, Changes, Tests and Experiments, and ITS 5.5.13, Technical Specifications (TS) Bases Control Program, are designed to assure that changes to the FSAR and ITS Bases do not result in changes to the Technical Specification requirements and do not result in significant increases in the probability or consequences of accidents previously evaluated, do not create the possibility of a new or different kind of accident, and do not result in a significant reduction in a margin of safety. Additionally, IP3 programs that implement FSAR changes in accordance with 10 CFR 50.59 and ITS Bases changes in accordance with ITS 5.5.13 require periodic submittal of FSAR and Bases changes to the NRC for review.



This change is a less restrictive administrative change with no impact on safety because no requirements are being deleted from Technical Specifications and an appropriate change control process and an appropriate level of regulatory oversight is maintained for the information being relocated out of the Technical Specifications.

LA.2 CTS 3.8.A.4 requires that the subcritical core be continuously monitored by two SRMs and that one SRM has audible indication in the containment available whenever core geometry is being changed. ITS LCO 3.9.1 maintains the requirement for two SRMs in Mode 6 and that Core Alterations and positive reactivity additions must be terminated if only one SRM is Operable: however, the detail related to audible indication in containment is not included in ITS LCO 3.9.2 and is relocated to the FSAR and will be implemented by plant procedures.

This change is acceptable because ITS 3.9.2 maintains the existing requirement for the Operability of two SRM channels (including visual indication and either alarms and/or audible indication in the control room). This ensures that redundant monitoring capability needed for the prompt identification of a boron dilution event is maintained in the Technical Specifications. However, audible SRM indication in the containment, which is intended to alert operators on the refueling floor of reactivity changes due to fuel mis-positioning, is not retained in Technical Specifications because ITS LCO 3.9.1, Boron Concentration, provides protection for this event.

This change is a less restrictive administrative change with no impact on safety because ITS 3.9.2 maintains the requirements to have audible SRM indication in the containment during changes in core geometry. Therefore; requirements to have audible SRM indication in the containment during changes in core geometry can be maintained in the TRM with no significant adverse impact on safety.



Conversion Package

Technical Specification 3.9.3: "Containmet Penetrations"

PART 1:



Containment Penetrations 3.9.3

3.9 REFUELING OPERATIONS

3.9.3 Containment Penetrations

LCO 3.9.3 The containment penetrations shall be in the following status:

a. The equipment hatch closed and held in place by at least four bolts or the equipment hatch opening is closed using an equipment hatch closure plate that may include a closed personnel access door;

b. One door in each air lock closed;

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1. closed by a manual or automatic isolation valve, a blind flange, or equivalent, or

2. capable of being closed by OPERABLE Containment Purge Isolation System.

---------------------------- NOTE ............................ LCO 3.9.3.d and LCO 3.9.3.e are not required to be met if the reactor has been subcritical for 2 550 hours. ..............-..............................-....-.........

d. The Containment Purge System flow path shall be either:

1. closed by a manual or automatic isolation valve, a blind flange, or equivalent, or

2. aligned to discharge through the HEPA filters and charcoal adsorbers.

e. The Containment Pressure Relief Line shall be closed by a manual or automatic isolation valve, a blind flange, or equivalent.

APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.

INDIAN POINT 3 Amendment [Rev.1j, 07/26/003.9.3-1


Conversion Package

Technical Specification 3.9.3: "Containment Penetrations"

PART 3:




DISCUSSION OF CHANGES ITS SECTION 3.9.3 - CONTAINMENT PENETRATIONS

aligned to discharge through the HEPA filters and charcoal adsorbers if fuel is being moved prior to a specified decay-time; however, ITS SR 3.9.3.2 requires verification every 7 days that this requirement is met. This change is needed because it requires periodic verification that LCO 3.9.3.d is being met. This change is acceptable because it does not introduce any operation that is un-analyzed while requiring periodic verification that LCO requirements are met. Therefore, this change has no adverse impact on safety.

M.3 CTS 3.8.A.9 requires that both the Containment Vent (i.e., 10 -inch pressure relief line per FSAR 5.3.2.4) and Containment Purge System (i.e., 36-inch supply and exhaust lines per FSAR 5.3.2.3) must be aligned to discharge through HEPA filters and charcoal adsorbers during movement of fuel until the reactor has been shutdown for a specified minimum number of hours.

ITS LCO 3.9.3.d maintains the requirement that the Containment Purge System (i.e., 36-inch supply and exhaust lines) be aligned to discharge through HEPA filters and charcoal adsorbers during movement of fuel until the reactor has been shutdown for a specified minimum number of hours; however, ITS LCO 3.9.3.e requires that the Containment Vent (i.e., 10 -inch pressure relief line) must be isolated until the reactor has been shutdown for a specified minimum number of hours. This change eliminates the option of having the Containment Vent (i.e., 10 -inch pressure relief line) open during the early stages of refueling.

This change is needed because the specific requirement in ITS LCO 3.9.3.e to keep the Containment Vent (i.e., 10 -inch pressure relief line) isolated until the reactor has been shutdown for a specified minimum number of hours is the justification for relocating requirements for testing of the associated HEPA filters and charcoal adsorbers from the Technical Specification to the Technical Requirements Manual (TRM). This change is acceptable because it is a reasonable interpretation of the existing requirement.

M.4 CTS 3.8.A.2 requires that at least one isolation valve shall be operable, locked closed or blind flanged in each line penetrating the containment. ITS 3.9.3.c.1 maintains this requirement except that, as clarified in the Bases, isolation may be accomplished by an OPERABLE automatic isolation valve. Specifying the operable isolation valve to be automatic is a more restricitve change. This change is needed to

Indian Point 3 5 ITS Conversion Submittal, Rev 1 '

DISCUSSION OF CHANGES ITS SECTION 3.9.3 - CONTAINMENT PENETRATIONS

provide assurance that this potential pathway for releases to the environment is closed in the event of a fuel-handling accident.

LESS RESTRICTIVE

L.1 CTS 3.8.A.8 requires that the Containment Building Vent and Purge System, including radiation monitors that initiate isolation, must be tested and verified to be operable within 100 hours prior to refueling operations. ITS SR 3.9.3.3 maintains the requirement for periodic verification that each required containment purge and exhaust valve actuates to the isolation position on an actual or simulated actuation signal; however, the SR Frequency is extended from within 100 hours prior to refueling operations to once every 92 days. This change is needed because it eliminates an ambiguity that could be interpreted as requiring the performance this SR every time refueling activities are started and stopped during a single refueling outage. This change is acceptable because the requirement that the Containment Building Vent and Purge isolation function is Operable during fuel movement and Core Alterations is unchanged. Elimination of the requirement to perform this verification within 100 hours prior to an activity is not significant because the normal periodic Surveillance Frequency is established to provide adequate assurance that requirements are being met. The 92-day Frequency ensures that the SR is performed at the start of each refueling and this Frequency is sufficient to provide a high degree of assurance that the valves will function as required throughout a refueling outage. Therefore, this change has no adverse impact on safety.

REMOVED DETAIL

None

Indian Point 3 6 TTq Cnnwarcinn kihm•t1• DU,, lvJ-~~~~ ~~ 11 -J .U I bl l l L III ba I, r-7 V IL


Conversion Package

Technical Specification 3.9.3: "Containment Penetrations"

PART 5:


NUREG-1431 and ITS


OG No. TSTF No. Generic Change Description NRC STATUS 1P3 STATUS JD No.

BWROG-017 051 RO REVISE CONTAJNMENT NRC Review Not Incorporated N/A REQUIREMENTS DURING HANDLING IRRADIATED FUEL AND CORE ALTERATIONS (REQUIREMENTS LIMITED TO "RECENTLY" IRRADIATED FUEL)

CEOG-021 068 Rx CONTAINMENT PERSONNEL See Next Rev. Superceded-See N/A - AIRLOCK DOORS OPEN DURING Next Rev

FUEL MOVEMENT

CEOG-021 R1 068 R1 CONTAINMENT PERSONNEL NRC Review Not Incorporated N/A AIRLOCK DOORS OPEN DURING FUEL MOVEMENT

CEOG-112 196 RO REVISE ISOLATION DEVICES TO Rejected by NRC Not Incorporated N/A INCLUDE ASME/ANSI EQUIVALENT METHODS



"Conversion Package

Technical Specification 3.9.3: " Containment Penetrations"

REQUIRE CONTAINMENT CLOSURE

REVISE THE CONTAINMENT PURGE AND EXHAUST SR TO EXEMPT VALVES THAT ARE LOCKED, SEALED OR SECURED

ADMINISTRATIVELY CONTROL CONTAINMENT PENETRATIONS

DEFINE OPERATION INVOLVING POSITIVE REATIVITY ADDITION

NRC Review

NRC Review

Rejected by TSTF

TSTF Review

Not Incorporated

Not Incorporated

Not Incorporated

Not Incorporated


CEOG-1 15

WOG-018

WOG-076

WOG-120

197 R1

092 R1

312 RO

286 RO

N/A

N/A

N/A

N/A

Containment Penetrations 3.9.P

3

3.9 REFUELING OPERATIONS

3.9.1 Containment Penetrations

4,2-S) LCO 3.9.#' The containment penetrations shall be in the following 3 status:

a. The equipment hatch closed and held in place by jfour.;

8 /..I b. One door in each air lock closed; *Rd

c. Each penetration providing direct access from the A,12.) containment atmosphere to the outside atmosphere either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

"2. capable of being closed by an OPERABLE Containment 02Purge a Isolation System.

APPLICABILITY:

ACTIONS

During CORE ALTERATIONS, During movement of irradiated fuel assemblies within

containment.

CONDITION REQUIRED ACTION COMPLETION TIME

A. One or more A.1 Suspend CORE Immediately containment ALTERATIONS. penetrations not in required status. AND

A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

WOG STS

6.K. V> ZI)OC A 8>

Rev 1, 04/07/95

NUREG-1431 Markup Inserts ITS SECTION 3.9.3 - CONTAINMENT PENETRATIONS

INSERT: 3.9-6-01

or the equipment hatch opening is closed using an equipment hatch closure plate that may include a closed personnel access door

INSERT: 3.9-6-02

- ---------------------------- NOTE ------------------------------LCO 3.9.3.d and LCO 3.9.3.e are not required to be met if the reactor has been subcritical for 2 550 hours.

ý a. S

<1Yx-ti$

d. The Containment Purge System flow path shall be either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. aligned to discharge through the HEPA filters and charcoal adsorbers.

e. The Containment Pressure Relief Line shall be closed by a manual or automatic isolation valve, a blind flange, or equivalent.

I e-

(9D

Containment Penetrations 3.9.1

FREQUENCY

7 days

)nU

aS

Rev 1, 04/07/95WOG STS 3.9-7


INSERT: 3.9-7-01

SR 3.9.3.2

DOC. -

---------------NOTE------------Not required to be met if the reactor has been subcritical for Ž 550 hours.

Verify Containment Purge System is either:

a. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

b. aligned to discharge through the HEPA filters and charcoal adsorbers.

7 days

j I______________________________

INSERT: 3.9-7-02

SR 3.9.3.4 -----------------.NOTE ------------------Not required to be met if the reactor has been subcritical for Ž 550 hours. - - - - - - - - - - - - - - - --------

Perform required Containment Purge System filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP

I ______________________________

Containment Penetrations B 3.9./

B 3.9 REFUELING OPERATIONS

B 3.9.# Containment Penetrations

BASES

BACKGROUND During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, a release of fission product radioactivity within containment will be restricted from escaping to the environment when the LCO requirements are met. In MODES 1, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1, "Containment." In MODE 6, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO requirements are referred to as "containment closure* rather than "containment OPERABILITY." Containment closure means that all potential escape paths are closed 4E7pibrioT>

g Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required.

The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, the equipment,,hatch must be held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced.

The containment air locks, which are also part of the 4` containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "Containment Air Locks." Each

- air lock has a door at both ends. The doors are-normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown

hOG STS Rev 1, 04/07/95

P


INSERT: B 3.9-11 - 01

In lieu of maintaining the equipment hatch inplace for containment (

closure, a temporary closure device may be used in l-cu of thc .quiprn.. t h-aee-to maintain containment closure during core alterations or during movement of irradiated fuel assemblies within containment. The temporary closure device may provide penetrations for temporary services or personnel access. The temporary closure device will be designed to withstand a seismic event and designed to a withstand a pressure which ensures containment closure during refueling operations. The closure device will provide the same level of protection as that of the equipment hatch for the fuel handling accident by restricting direct air flow from the containment to the environment.

Containment Penetrations B 3.9.!f

.3

BASES

BACKGROUND (continuel

1AXvQA.

when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, containment closure is required; therefore, the door interlock mechanism may remain disabled, but one air lock door must always remain closed.

The requirements for contairmnent penetration closure ensure that a release of fission product radioactivity within containment will be restricted from escaping to the environment. The closure restrictions are sufficient to restrict fission product radioactivity release from containment due to a fuel handling accident during refueling.

The Con inment Purge and Lxhaust System includes two subs., ems. The normal usystem includes a .4 inch purgee pe tration and a 42 ijýoh exhaust penetrato. The second

•bsystem, a minipu2 e system, includes a• 8 inch purge

penetration and. 8 inch exhaust penetration. During NODES 1, 2, 3, nd 4, the two valvee'in each of the-normal purge and ex ust penetrations ar.'secured in the clofid position. he two valves in jeach of the two minip r~e penetra ons can be opened/ntermittently, but =' closed aut ically by the Eng~ieered Safety FeaturpeActuation Sy em (ESFAS). Neithe' of the subsystems subject to a

ecification in NODE5.

In NODE 6, larlg air exchange~s ar necessary to conduct ,

refueling op6rations. The cnormal !2 inch purge system lz used forithis purpose, and ajPl'four valves are closeIy the ESFASJi accordance with LCO 3.3.2, "Engineered Saftty Feature Actuation System•(ESFAS) Instrumentatiom<."

"-"The minipurge system remains operationa"n ODE 6, and all rfour valves are ,also closed by the ESFAS. S/"orX

LThe minipurge system is no,(used in MODE 6. All four"8 inch valves are secured in the closed position.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere

I" e et4dttuff

d)

Rev 1, 04/07/95WOG STS B 3.9-12


INSERT: B 3.9-12-01 &.c

The Containment Purge System consists of~a-36-inch containment purge supply and exhaust. The supply system includes roughing filters, heating coils, fan and a containment penetration with two butterfly valves for isolation. The exhaust system includes a containment penetration with two butterfly valves for isolation and can be aligned to discharge to the atmosphere through the plant vent either directly or through the Containment Purge Filter System (i.e., a filter bank with roughing, HEPA and charcoal filters).

The Containment Purge System must be isolated when in Modes 1, 2, 3 or 4 in accordance with requirements established in LCO 3.6.3, Containment Isolation Valves. In Modes 5 and 6, the Containment Purge System may be used for containment ventilation. When open, the Containment Purge System isolation valves are capable of closing in response to the detection of high radiation levels in accordance with requirements established in LCO 3.3.6, Containment Purge and Pressure Relief Isolation Instrumentation (Ref. 1). Despite this isolation capability, the Containment Purge System must be aligned to discharge through the Containment Purge Filter System during CORE ALTERATIONS or movement of irradiated fuel until the reactor has been shutdown for a specified minimum number of hours.


INSERT: B 3.9-12-01 (continued)

The Containment Pressure Relief Line (i.e., Containment Vent) consists of a single 10-inch containment vent line that is used to handle normal pressure changes in the Containment when in Modes 1, 2, 3 and 4 (Ref. 1). The Containment Pressure Relief Line is equipped with three quick-closing butterfly type isolation valves, one inside and two outside the containment which isolate automatically in accordance with requirements established in LCO 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, and LCO 3.3.6, Containment Purge and Pressure Relief Isolation Instrumentation. Although the Containment Pressure Relief Line discharges to the atmosphere via the Containment Auxiliary Charcoal Filter System (i.e., a filter bank with roughing, HEPA and charcoal filters), the Containment Pressure Relief Line must remain isolated during CORE ALTERATIONS or movement of irradiated fuel until the reactor has been shutdown for a specified minimum number of hours. The Containment Pressure Relief Line must remain isolated because the Containment Auxiliary Charcoal Filter System is not required to be tested in accordance with Specification 5.5.10, Ventilation Filter Test Program,

Containment Penetrations B 3.9.,K (

BASES

BACKGROUND must be isolated on at least one side. Isolation may be (continued) achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or esuivalent. -/YLEquiva en iso a ion methods must be approvedand may

(Ayito 1 Fe sosi include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during fuel movements ARef-.+).

APPLICABLE During CORE ALTERATIONS-or movement of irradiated fuel SAFETY ANALYSES assemblies within containment, the most severe radiological consequences result from a fuel handling accident. The fuel handling accident is a postulated event that involves damage to irradiated fuel (Ref. 2). Fuel handling accidents, analyzed in leference(-, include dropping a single irradiated fuel assembly and handling tool or a heavy object " nto other irradiated fuel assemblies erequirements M

LCO 3ý, 7, '"efuel fg ca ýity Wagr Le 0, 0 ndW ih•e u• eca4 time 4f 100/hours nrior ACOR[AL RATIO

FTrA. that the release of fission product radioactivity,subsequent to a fuel handling accident, results in doses that are well within the guideline values specified in 10 CFR 100. Standard Review Plan, Section 15.7.4, Rev. 1 (Ref. 3), defines "well within' 10 CFR 100 to be 25% or less of the 10 CFR 100 values. The acceptance limits for offsite radiation exposure will be 25% of 10 CFR 100 values or the NRC staff approved licensing basis (e.g., a specified fraction of 10 CFR 100 limits).

Containment penetrations satisfy Criterion 3 of (0 To iMz; ie en-.

LCO This LCO limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment. The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmospher, to e closed except for the OPERABLE containment purge cg

etrations. For the OPERABLE containment purge "~ penetrations, this LCO ensures that these peeptrations are isolable by the Containment Purge and U _Xf Isolation G_4W., The OPERABILITY requirements for 'M- ensure tha• eautomatic purge n valves

-01Z -feanti-nued)

Rev 1, 04/07/95MOG STS B 3.9-13


INSERT: B 3.9-13-01

The release of radioactivity from the containment following a fuel handling accident is limited by the following:

a) The requirements of LCO 3.9.6, "Refueling Cavity Water Level;"

b) The minimum decay time of 145 hours prior to CORE ALTERATIONS; and,

c) The requirements of this LCO to either isolate the Containment Purge System or align the system to discharge through the HEPA filters and charcoal adsorbers for a minimum of first 550 hours following the reactor shutdown.

This combination of requirements ensures

INSERT: B 3.9-13-02

Additionally, the requirement to isolate the Containment Purge System orC align the system to discharge through the HEPA filters and charcoal adsorbers for a minimum of the first 550 hours following the reactor shutdown is required to limit offsite radiation exposure to within required limits. The Containment Pressure Relief Line must remain isolated because the Containment Auxiliary Charcoal Filter System is not required to be tested in accordance with Specification 5.5.10, Ventilation Filter Test Program,

Containment Penetrations B .9 '

BASES

LCO loss i nt / - a. (continued) umpefie t te ssu mptions used in the safety analysis

to ensure that releases through the valves terminated, suc a ra iologica doses are within the acceptance limit.

APPLICABILITY

ACTIONS

The containment penetration requirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when there is a potential for a fuel han.dling accident. In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1. In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status.

A.1 and A.2

If the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, including the Containment Purge and

s-Isolation not capable of automatic actuation D ýwhen the purge and exhaust valves are open, the unit must be

placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9..L REQUIREMENTS "R E This Surveillance demonstrates that each of the containmLnt

penetrations required to be in its closed position is in that position. The Surveillance on the open purge and exhaust valves will demonstrate that the valves are not blocked from closing. Also the Surveillance will

t I

!

8 3.9-14 Rev 1, 04/07/95WOG ST"S

Containment Penetrations B 3.9.X

BASES

SURVFT !aurr - - -REQUIREMENTS Sj (continued)

demonstrate that each valve operator has motive Power, which will ensure that each valve is capable of being closed by an OPERABLE automatic containment purge and exhaust isolation signal.

The Surveillance is performed every 7 days during CORE ALTERATIONS or movement of Arradiated fuel assemblies within containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations. A surveillance before the start of refueling operations will provide two or three surveillance verifications during the applicable period for this LCO. As such, this Surveillance ensures that a postulated fuel handling accident that releases fission product radioactivity within the containment will not result in a release of fission product radioactivity to the environment.

This Surveillance demonstrates that each containment purge and exhaust valve actuates to its isolation •an cuao position on m a c lor simulated high radiation signaj .~ 0 -

'o nrainment "-s J iolnt~o~ instrumentation requires a CHANNEL CHECK every 12 hours and a COT every 92 days to ensure the channel OPERABILITY during refueling operations. Lvery months a C HANEL C ALIB I is performed. fThVr f Y sW M 1tama ~ u ~ ~ l \ 0e~ n R T O _

--- aj emonstrates thait te isolation ti'me ofeac valve is in accordance w ie Inservice Testing Program requirements. These Surveillances Performed during MODE 6 will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment. S

4,

VOG STS 40e. B 3.9-15

Rev 1, 04/07/95

-i�J..

Tr�AA'

639- i5�-o�


INSERT: B 3.9-15-01

SR 3.9.3.2

This SR requires periodic verification every 7 days that the Containment Building Purge System is either isolated or aligned to discharge through the HEPA filters and charcoal adsorbers. This SR is needed because it requires periodic verification that LCO 3.9.3.d is being met. A Note provides the allowance that this SR is not required to be performed or met if the reactor has been subcritical for Ž 550 hours. These restrictions ensure that the offsite dose limit for a fuel handling accident of 75 rem to the thyroid at the exclusion area boundary (i.e., 25 percent of the 10 CFR Part 100 limit of 300 rem) is met by either filtering any release from the containment or by allowing a greater decay time before fuel handling activities are permitted.

INSERT: B 3.9-15-02

ensures that this SR is performed prior to this function being required and periodically thereafter.

INSERT: B 3.9-15-03

SR 3.9.3.4

This SR verifies that the required Containment Building Purge System testing is performed in accordance with Specification 5.5.10, Ventilation Filter Test Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP.

Containment Penetrations B 3.9..

BASES (continued)

REFERENCES 1. luat-

2. FSAR, Section M1ij. j

3. NUREG-0800, Section 15.7.4, Rev. 1, July 1981.

Rev 1, 04/07/95WOG STS B 3.9-16


Conversion Package

Technical Specification 3.9.5: "Residual Heat Removal (RHR) and Coolant

Circulation - Low Water Level"

PART 1:


Rl1-V

UICv, o


RHR and Cool ant Circulation - Low Water Level B 3.9.5

BASES


SR 3_.9_Li (continued)

water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The Frequency of 12 hours is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator for monitoring the RHR System in the control room.

Verification that the required pump not in operation is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.


INDIAN POINT 3 Revision [Rev.1], 07/26/00B 3.9.5- 4

Documents

Technical Specification 3.7.10, Ultimate Heat Sink (UHS)